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Sample records for graphite irradiation creep

  1. Irradiation Creep in Graphite

    Energy Technology Data Exchange (ETDEWEB)

    Ubic, Rick; Butt, Darryl; Windes, William

    2014-03-13

    An understanding of the underlying mechanisms of irradiation creep in graphite material is required to correctly interpret experimental data, explain micromechanical modeling results, and predict whole-core behavior. This project will focus on experimental microscopic data to demonstrate the mechanism of irradiation creep. High-resolution transmission electron microscopy should be able to image both the dislocations in graphite and the irradiation-induced interstitial clusters that pin those dislocations. The team will first prepare and characterize nanoscale samples of virgin nuclear graphite in a transmission electron microscope. Additional samples will be irradiated to varying degrees at the Advanced Test Reactor (ATR) facility and similarly characterized. Researchers will record microstructures and crystal defects and suggest a mechanism for irradiation creep based on the results. In addition, the purchase of a tensile holder for a transmission electron microscope will allow, for the first time, in situ observation of creep behavior on the microstructure and crystallographic defects.

  2. Significance of primary irradiation creep in graphite

    CSIR Research Space (South Africa)

    Erasmus, C

    2013-05-01

    Full Text Available Traditionally primary irradiation creep is introduced into graphite analysis by applying the appropriate amount of creep strain to the model at the initial time-step. This is valid for graphite components that are subjected to high fast neutron flux...

  3. An analysis of irradiation creep in nuclear graphites

    International Nuclear Information System (INIS)

    Neighbour, G.B.; Hacker, P.J.

    2002-01-01

    Nuclear graphite under load shows remarkably high creep ductility with neutron irradiation, well in excess of any strain experienced in un-irradiated graphite (and additional to any dimensional changes that would occur without stress). As this behaviour compensates, to some extent, some other irradiation effects such as thermal shutdown stresses, it is an important property. This paper briefly reviews the approach to irradiation creep in the UK, described by the UK Creep Law. It then offers an alternative analysis of irradiation creep applicable to most situations, including HTR systems, using AGR moderator graphite as an example, to high values of neutron fluence, applied stress and radiolytic weight loss. (authors)

  4. Irradiation creep performance of graphite relevant for pebble bed HTRs

    International Nuclear Information System (INIS)

    Kleist, G.; O'Connor, M.F.

    1980-01-01

    Irradiation - induced creep in the core reflector component graphite of high temperature reactors is of primary importance to the core designer since it provides a mechanism for the relief of internal stresses arising from differential Wigner shrinkage and thermal expansion. The experimental determination of the extent of this creep for conditions relevant to the reactor is thus imperative

  5. Irradiation creep in reactor graphites for HTR applications. [Neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Veringa, H J; Blackstone, R [Stichting Reactor Centrum Nederland, Petten

    1976-01-01

    A series of restrained shrinkage experiments on a number of graphites in the temperature range 400 to 1400/sup 0/C is described. A description is given of the experimental method and method of data evaluation. The results are compared with data from other sources. Analysis of data confirms that the creep coefficient, which is defined as the radiation induced creep strain per unit stress per unit neutron fluence, is inversely proportional to the pre-irradiation value of the Young's modulus of the material. The radiation creep coefficient increases with temperature in the range 400 to 1400/sup 0/C. It can be represented by the sum of two temperature dependent functions, one of which is inversely proportional to the neutron flux density, the other independent of the neutron flux density. When the data are analysed in this way it is found that the graphites investigated in the present work, although made from widely different starting materials and by different processes, show the same dependence of the irradiation creep coefficient on the temperature and the neutron flux density.

  6. Irradiation-induced creep in graphite: a review

    International Nuclear Information System (INIS)

    Price, R.J.

    1981-08-01

    Data on irradiation-induced creep in graphite published since 1972 are reviewed. Sources include restrained shrinkage tests conducted at Petten, the Netherlands, tensile creep experiments with continuous strain registration at Petten and Grenoble, France, and controlled load tests with out-of-reactor strain measurement performed at Oak Ridge National Laboratory, Petten, and in the United Kingdom. The data provide reasonable confirmation of the linear viscoelastic creep model with a recoverable transient strain component followed by a steady-state strain component, except that the steady-state creep coefficient must be treated as a function of neutron fluence and is higher for tensile loading than for compressive loading. The total transient creep strain is approximately equal to the preceding elastic strain. No temperature dependence of the transient creep parameters has been demonstrated. The initial steady-state creep coefficient is inversely proportional to the unirradiated Young modulus

  7. Examination of Experimental Data for Irradiation - Creep in Nuclear Graphite

    Science.gov (United States)

    Mobasheran, Amir Sassan

    The objective of this dissertation was to establish credibility and confidence levels of the observed behavior of nuclear graphite in neutron irradiation environment. Available experimental data associated with the OC-series irradiation -induced creep experiments performed at the Oak Ridge National Laboratory (ORNL) were examined. Pre- and postirradiation measurement data were studied considering "linear" and "nonlinear" creep models. The nonlinear creep model considers the creep coefficient to vary with neutron fluence due to the densification of graphite with neutron irradiation. Within the range of neutron fluence involved (up to 0.53 times 10^{26} neutrons/m ^2, E > 50 KeV), both models were capable of explaining about 96% and 80% of the variation of the irradiation-induced creep strain with neutron fluence at temperatures of 600^circC and 900^circC, respectively. Temperature and reactor power data were analyzed to determine the best estimates for the actual irradiation temperatures. It was determined according to thermocouple readouts that the best estimate values for the irradiation temperatures were well within +/-10 ^circC of the design temperatures of 600^circC and 900 ^circC. The dependence of the secondary creep coefficients (for both linear and nonlinear models) on irradiation temperature was determined assuming that the variation of creep coefficient with temperature, in the temperature range studied, is reasonably linear. It was concluded that the variability in estimate of the creep coefficients is definitely not the results of temperature fluctuations in the experiment. The coefficients for the constitutive equation describing the overall growth of grade H-451 graphite were also studied. It was revealed that the modulus of elasticity and the shear modulus are not affected by creep and that the electrical resistivity is slightly (less than 5%) changed by creep. However, the coefficient of thermal expansion does change with creep. The consistency of

  8. The irradiation creep characteristics of graphite to high fluences

    International Nuclear Information System (INIS)

    Kennedy, C.R.; Cundy, M.; Kleist, G.

    1988-01-01

    High-temperature gas-cooled reactors (HTGR) have massive blocks of graphite with thermal and neutron-flux gradients causing high internal stresses. Thermal stresses are transient; however, stresses generated by differential growth due to neutron damage continue to increase with time. Fortunately, graphite also experiences creep under irradiation allowing relaxation of stresses to nominally safe levels. Because of complexity of irradiation creep experiments, data demonstrating this phenomenon are generally limited to fairly low fluences compared to the overall fluences expected in most reactors. Notable exceptions have been experiments at 300/degree/C and 500/degree/C run at Petten under tension and compression creep stresses to fluences greater than 4 /times/ 10 26 (E > 50 keV) neutrons/m 2 . This study complements the previous results by extending the irradiation temperature to 900/degree/C. 2 refs., 3 figs

  9. The irradiation creep in reactor graphites for HTR applications

    International Nuclear Information System (INIS)

    Veringa, H.J.; Blackstone, R.

    1976-01-01

    A series of restrained shrinkage experiments on a number of graphites in the temperature range 400 to 1400 0 C is described. A description is given of the experimental method and method of data evaluation. The results are compared with data from other sources. Analysis of data confirms that the creep coefficient, which is defined as the radiation induced creep strain per unit stress per unit neutron fluence, is inversely proportional to the pre-irradiation value of the Young's modulus of the material. The radiation creep coefficient increases with temperature in the range 400 to 1400 0 C. It can be represented by the sum of two temperature dependent functions, one of which is inversely proportional to the neutron flux density, the other independent of the neutron flux density. When the data are analysed in this way it is found that the graphites investigated in the present work, although made from widely different starting materials and by different processes, show the same dependence of the irradiation creep coefficient on the temperature and the neutron flux density. (author)

  10. Project accent: graphite irradiated creep in a materials test reactor

    International Nuclear Information System (INIS)

    Brooking, M.

    2014-01-01

    Atkins manages a pioneering programme of irradiation experiments for EDF Energy. One of these projects is Project ACCENT, designed to obtain evidence of a beneficial physical property of the graphite, which may extend the life of the Advanced Gas-cooled Reactors (AGRs). The project team combines the in-house experience of EDF Energy with two supplier organisations (providing the material test reactors and testing facilities) and supporting consultancies (Atkins and an independent technical expert). This paper describes: - Brief summary of the Project; - Discussion of the challenges faced by the Project; and - Conclusion elaborating on the aims of the Project. These challenging experiments use bespoke technology and both un-irradiated (virgin) and irradiated AGR graphite. The results will help to better understand graphite irradiation-induced creep (or stress modified dimensional change) properties and therefore more accurately determine lifetime and safe operating envelopes of the AGRs. The first round of irradiation has been completed, with a second round about to commence. This is a key step to realising the full lifetime ambition for AGRs, demonstrating the relaxation of stresses within the graphite bricks. (authors)

  11. The irradiation induced creep of graphite under accelerated damage produced by boron doping

    International Nuclear Information System (INIS)

    Brocklehurst, J.E.

    1975-01-01

    The presence of boron enhances fast neutron irradiation damage in graphite by providing nucleation sites for interstitial loop formation. Doping with 11 B casues an increase in the irradiation induced macroscopic dimensional changes, which have been shown to result from an acceleration in the differential crystal growth rate for a given carbon atom displacement rate. Models of irradiation induced creep in graphite have centred around those in which creep is induced by internal stresses due to the anisotopic crystal growth, and those in which creep is activated by atomic displacements. A creep test on boron doped graphite has been performed in an attempt to establish which of these mechanisms is the determining factor. An isotropic nuclear graphite was doped to a 11 B concentration of 0.27 wt.%. The irradiation induced volume shrinkage rate at 750 0 C increased by a factor of 3 over that of the virgin graphite, in agreement with predictions from the earlier work, but the total creep strains were comparable in both doped and virgin samples. This observation supports the view that irradiation induced creep is dependent only on the carbon atom displacement rate and not on the internal stress level determined by the differential crystal growth rate. The implications of this result on the irradiation behaviour of graphite containing significant concentrations of boron are briefly discussed. (author)

  12. High temperature graphite irradiation creep experiment in the Dragon Reactor. Dragon Project report

    Energy Technology Data Exchange (ETDEWEB)

    Manzel, R.; Everett, M. R.; Graham, L. W.

    1971-05-15

    The irradiation induced creep of pressed Gilsocarbon graphite under constant tensile stress has been investigated in an experiment carried out in FE 317 of the OECD High Temperature Gass Cooled Reactor ''Dragon'' at Winfrith (England). The experiment covered a temperature range of 850 dec C to 1240 deg C and reached a maximum fast neutron dose of 1.19 x 1021 n cm-2 NDE (Nickel Dose DIDO Equivalent). Irradiation induced dimensional changes of a string of unrestrained graphite specimens are compared with the dimensional changes of three strings of restrained graphite specimens stressed to 40%, 58%, and 70% of the initial ultimate tensile strength of pressed Gilsocarbon graphite. Total creep strains ranging from 0.18% to 1.25% have been measured and a linear dependence of creep strain on applied stress was observed. Mechanical property measurements carried out before and after irradiation demonstrate that Gilsocarbon graphite can accommodate significant creep strains without failure or structural deterioration. Total creep strains are in excellent agreement with other data, however the results indicate a relatively large temperature dependent primary creep component which at 1200 deg C approaches a value which is three times larger than the normally assumed initial elastic strain. Secondary creep constants derived from the experiment show a temperature dependence and are in fair agreement with data reported elsewhere. A possible determination of the results is given.

  13. Experimental Plan for EDF Energy Creep Rabbit Graphite Irradiations- Rev. 2 (replaces Rev. 0 ORNL/TM/2013/49).

    Energy Technology Data Exchange (ETDEWEB)

    Burchell, Timothy D [ORNL

    2014-07-01

    The experimental results obtained here will assist in the development and validation of future models of irradiation induced creep of graphite by providing the following data: Inert creep stain data from low to lifetime AGR fluence Inert creep-property data (especially CTE) from low to lifetime AGR fluence Effect of oxidation on creep modulus (by indirect comparison with experiment 1 and direct comparison with experiment 3 NB. Experiment 1 and 3 are not covered here) Data to develop a mechanistic understanding, including oAppropriate creep modulus (including pinning and high dose effects on structure) oInvestigation of CTE-creep strain behavior under inert conditions oInformation on the effect of applied stress/creep strain on crystallite orientation (requires XRD) oEffect of creep strain on micro-porosity (requires tomography & microscopy) This document describes the experimental work planned to meet the requirements of project technical specification [1] and EDF Energy requests for additional Pre-IE work. The PIE work is described in detail in this revision (Section 8 and 9).

  14. Irradiation induced creep in graphite with respect to the flux effect and the high fluence behaviour

    International Nuclear Information System (INIS)

    Cundy, M.R.

    1984-01-01

    In accelerated irradiation creep tests, performed in the HFR Petten, in a fast neutron flux of about 2x10 4 cm -2 s -1 and at temperatures of 300 and 500 0 C, a fast neutron fluence in excess of 20x10 21 cm -2 (EDN) has been attained so far. As a supplement to this, an analogous creep test was conducted in a fast neutron flux lower by a factor of four which is more typical for the service conditions in a HTR, with a maximum fast fluence of only 4x10 21 cm -2 (EDN). This experiment was aimed at answering the question if, for equal fast fluence, enhanced irradiation creep and Wigner dimensional change would take place in a reduced fast neutron flux. This problem has more generally been addressed to as the ''flux effect'' or the ''equivalent temperature concept''. (orig./IHOE)

  15. Radiation creep of graphite. An introduction

    Energy Technology Data Exchange (ETDEWEB)

    Blackstone, R [Commission of the European Communities, Petten (Netherlands). Joint Nuclear Research Center

    1977-03-01

    Graphite, a class of materials with many unique and unusual properties, shows a remarkably high creep ductility under irradiation. As this behaviour compensates to some extent some of the more worrying radiation effects, such as dimensional changes and their strong temperature dependence, it is a property of large technological interest. There are various ways of observing and measuring in-pile creep of graphite, varying in degree of sophistication and in cost, in accuracy and in the type of data that is generated. This paper attempts to review briefly the various experimental methods, and the knowledge generated so far. An indication is given of the areas in which further knowledge is wanted.

  16. Radiation creep of graphite. An introduction

    International Nuclear Information System (INIS)

    Blackstone, R.

    1977-01-01

    Graphite, a class of materials with many unique and unusual properties, shows a remarkably high creep ductility under irradiation. As this behavior compensates to some extent some of the more worrying radiation effects, such as dimensional changes and their strong temperature dependence, it is a property of large technological interest. There are various ways of observing and measuring in-pile creep of graphite, varying in degree of sophistication and in cost, in accuracy and in the type of data that is generated. This paper attempts to review briefly the various experimental methods, and the knowledge generated so far. An indication is given of the areas in which further knowledge is wanted

  17. Radiation creep of graphite. An introduction

    International Nuclear Information System (INIS)

    Blackstone, R.

    1977-01-01

    Graphite, a class of materials with many unique and unusual properties, shows a remarkably high creep ductility under irradiation. As this behaviour compensates to some extent some of the more worrying radiation effects, such as dimensional changes and their strong temperature dependence, it is a property of large technological interest. There are various ways of observing and measuring in-pile creep of graphite, varying in degree of sophistication and in cost, in accuracy and in the type of data that is generated. This paper attempts to review briefly the various experimental methods, and the knowledge generated so far. An indication is given of the areas in which further knowledge is wanted. (Auth.)

  18. Irradiation creep models - an overview

    International Nuclear Information System (INIS)

    Matthews, J.R.; Finnis, M.W.

    1988-01-01

    The modelling of irradiation creep is now highly developed but many of the basic processes underlying the models are poorly understood. A brief introduction is given to the theory of cascade interactions, point defect clustering and dislocation climb. The range of simple irradiation creep models is reviewed including: preferred nucleation of interstitial loops; preferred absorption of point defects by dislocations favourably orientated to an applied stress; various climb-enhanced glide and recovery mechanisms, and creep driven by internal stresses produced by irradiation growth. A range of special topics is discussed including: cascade effects; creep transients; structural and induced anisotropy; and the effect of impurities. The interplay between swelling and growth with thermal and irradiation creep is emphasized. A discussion is given on how irradiation creep theory should best be developed to assist the interpretation of irradiation creep observations and the requirements of reactor designers. (orig.)

  19. AGC 2 Irradiation Creep Strain Data Analysis

    International Nuclear Information System (INIS)

    Windes, William E.; Rohrbaugh, David T.; Swank, W. David

    2016-01-01

    The Advanced Reactor Technologies Graphite Research and Development Program is conducting an extensive graphite irradiation experiment to provide data for licensing of a high temperature reactor (HTR) design. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor designs. Nuclear graphite H-451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphite grades have been developed and are considered suitable candidates for new HTR reactor designs. To support the design and licensing of HTR core components within a commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance are required for the physical, mechanical, and thermal properties of each graphite grade, with a specific emphasis on data accounting for the life limiting effects of irradiation creep on key physical properties of the HTR candidate graphite grades. Further details on the research and development activities and associated rationale required to qualify nuclear grade graphite for use within the HTR are documented in the graphite technology research and development plan.

  20. AGC 2 Irradiation Creep Strain Data Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Windes, William E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rohrbaugh, David T. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Swank, W. David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-08-01

    The Advanced Reactor Technologies Graphite Research and Development Program is conducting an extensive graphite irradiation experiment to provide data for licensing of a high temperature reactor (HTR) design. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor designs. Nuclear graphite H-451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphite grades have been developed and are considered suitable candidates for new HTR reactor designs. To support the design and licensing of HTR core components within a commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance are required for the physical, mechanical, and thermal properties of each graphite grade, with a specific emphasis on data accounting for the life limiting effects of irradiation creep on key physical properties of the HTR candidate graphite grades. Further details on the research and development activities and associated rationale required to qualify nuclear grade graphite for use within the HTR are documented in the graphite technology research and development plan.

  1. Simulation of irradiation creep

    International Nuclear Information System (INIS)

    Reiley, T.C.; Jung, P.

    1977-01-01

    The results to date in the area of radiation enhanced deformation using beams of light ions to simulate fast neutron displacement damage are reviewed. A comparison is made between these results and those of in-reactor experiments. Particular attention is given to the displacement rate calculations for light ions and the electronic energy losses and their effect on the displacement cross section. Differences in the displacement processes for light ions and neutrons which may effect the irradiation creep process are discussed. The experimental constraints and potential problem areas associated with these experiments are compared to the advantages of simulation. Support experiments on the effect of thickness on thermal creep are presented. A brief description of the experiments in progress is presented for the following laboratories: HEDL, NRL, ORNL, PNL, U. of Lowell/MIT in the United States, AERE Harwell in the United Kingdom, CEN Saclay in France, GRK Karlsruhe and KFA Julich in West Germany

  2. Intercomparison of graphite irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Hering, H; Perio, P; Seguin, M [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1959-07-01

    While fast neutrons only are effective in damaging graphite, results of irradiations are more or less universally expressed in terms of thermal neutron fluxes. This paper attempts to correlate irradiations made in different reactors, i.e., in fluxes of different spectral compositions. Those attempts are based on comparison of 1) bulk length change and volume expansion, and 2) crystalline properties (e.g., lattice parameter C, magnetic susceptibility, stored energy, etc.). The methods used by various authors for determining the lattice constants of irradiated graphite are discussed. (author)

  3. Measuring irradiation creep

    International Nuclear Information System (INIS)

    Pelah, I.

    1981-03-01

    Simulation of fusion-neutron induced damage by beams of light ions is discussed. It is suggested that accelerated creep measurements to determine ''end of life'' of materials may be done by the application of thermal treatment and thermal creep measurements. (author)

  4. Neutron irradiation creep in stainless steel alloys

    Energy Technology Data Exchange (ETDEWEB)

    Schuele, Wolfgang (Commission of the European Union, Institute for Advanced Materials, I-21020 Ispra (Vatican City State, Holy See) (Italy)); Hausen, Hermann (Commission of the European Union, Institute for Advanced Materials, I-21020 Ispra (Vatican City State, Holy See) (Italy))

    1994-09-01

    Irradiation creep elongations were measured in the HFR at Petten on AMCR steels, on 316 CE-reference steels, and on US-316 and US-PCA steels varying the irradiation temperature between 300 C and 500 C and the stress between 25 and 300 MPa. At the beginning of an irradiation a type of primary'' creep stage is observed for doses up to 3-5 dpa after which dose the secondary'' creep stage begins. The primary'' creep strain decreases in cold-worked steel materials with decreasing stress and decreasing irradiation temperature achieving also negative creep strains depending also on the pre-treatment of the materials. These primary'' creep strains are mainly attributed to volume changes due to the formation of radiation-induced phases, e.g. to the formation of [alpha]-ferrite below about 400 C and of carbides below about 700 C, and not to irradiation creep. The secondary'' creep stage is found for doses larger than 3 to 5 dpa and is attributed mainly to irradiation creep. The irradiation creep rate is almost independent of the irradiation temperature (Q[sub irr]=0.132 eV) and linearly dependent on the stress. The total creep elongations normalized to about 8 dpa are equal for almost every type of steel irradiated in the HFR at Petten or in ORR or in EBR II. The negative creep elongations are more pronounced in PCA- and in AMCR-steels and for this reason the total creep elongation is slightly smaller at 8 dpa for these two steels than for the other steels. ((orig.))

  5. Irradiation creep in simple binary alloys

    International Nuclear Information System (INIS)

    Nagakawa, J.; Sethi, V.K.; Turner, A.P.L.

    1981-07-01

    Creep enhancement during 21-MeV deuteron irradiation was examined at 350 0 C for two simple binary alloys with representative microstructures, i.e., solid-solution (Ni - 4 at. % Si) and precipitation-hardened (Ni - 12.8 at. % Al) alloys. Coherent precipitates were found to be very effective in suppressing irradiation-enhanced creep. Si solute atoms depressed irradiation creep moderately and caused irradiation hardening via radiation-induced segregation. The stress-dependence of irradiation creep in Ni - 4 at. % Si should a transition, which seems to reflect a change of mechanism from dislocation climb due to stress-induced preferential absorption (SIPA) to climb-controlled dislocation glide enhanced by irradiation

  6. Irradiation creep of dispersion strengthened copper alloy

    International Nuclear Information System (INIS)

    Pokrovsky, A.S.; Barabash, V.R.; Fabritsiev, S.A.

    1997-01-01

    Dispersion strengthened copper alloys are under consideration as reference materials for the ITER plasma facing components. Irradiation creep is one of the parameters which must be assessed because of its importance for the lifetime prediction of these components. In this study the irradiation creep of a dispersion strengthened copper (DS) alloy has been investigated. The alloy selected for evaluation, MAGT-0.2, which contains 0.2 wt.% Al 2 O 3 , is very similar to the GlidCop trademark alloy referred to as Al20. Irradiation creep was investigated using HE pressurized tubes. The tubes were machined from rod stock, then stainless steel caps were brazed onto the end of each tube. The creep specimens were pressurized by use of ultra-pure He and the stainless steel caps subsequently sealed by laser welding. These specimens were irradiated in reactor water in the core position of the SM-2 reactors to a fluence level of 4.5-7.1 x 10 21 n/cm 2 (E>0.1 MeV), which corresponds to ∼3-5 dpa. The irradiation temperature ranged from 60-90 degrees C, which yielded calculated hoop stresses from 39-117 MPa. A mechanical micrometer system was used to measure the outer diameter of the specimens before and after irradiation, with an accuracy of ±0.001 mm. The irradiation creep was calculated based on the change in the diameter. Comparison of pre- and post-irradiation diameter measurements indicates that irradiation induced creep is indeed observed in this alloy at low temperatures, with a creep rate as high as ∼2 x 10 -9 s -1 . These results are compared with available data for irradiation creep for stainless steels, pure copper, and for thermal creep of copper alloys

  7. Irradiation creep of dispersion strengthened copper alloy

    Energy Technology Data Exchange (ETDEWEB)

    Pokrovsky, A.S.; Barabash, V.R.; Fabritsiev, S.A. [and others

    1997-04-01

    Dispersion strengthened copper alloys are under consideration as reference materials for the ITER plasma facing components. Irradiation creep is one of the parameters which must be assessed because of its importance for the lifetime prediction of these components. In this study the irradiation creep of a dispersion strengthened copper (DS) alloy has been investigated. The alloy selected for evaluation, MAGT-0.2, which contains 0.2 wt.% Al{sub 2}O{sub 3}, is very similar to the GlidCop{trademark} alloy referred to as Al20. Irradiation creep was investigated using HE pressurized tubes. The tubes were machined from rod stock, then stainless steel caps were brazed onto the end of each tube. The creep specimens were pressurized by use of ultra-pure He and the stainless steel caps subsequently sealed by laser welding. These specimens were irradiated in reactor water in the core position of the SM-2 reactors to a fluence level of 4.5-7.1 x 10{sup 21} n/cm{sup 2} (E>0.1 MeV), which corresponds to {approx}3-5 dpa. The irradiation temperature ranged from 60-90{degrees}C, which yielded calculated hoop stresses from 39-117 MPa. A mechanical micrometer system was used to measure the outer diameter of the specimens before and after irradiation, with an accuracy of {+-}0.001 mm. The irradiation creep was calculated based on the change in the diameter. Comparison of pre- and post-irradiation diameter measurements indicates that irradiation induced creep is indeed observed in this alloy at low temperatures, with a creep rate as high as {approx}2 x 10{sup {minus}9}s{sup {minus}1}. These results are compared with available data for irradiation creep for stainless steels, pure copper, and for thermal creep of copper alloys.

  8. Review of recent irradiation-creep results

    International Nuclear Information System (INIS)

    Coghlan, W.A.

    1982-05-01

    Materials deform faster under stress in the presence of irradiation by a process known as irradiation creep. This phenomenon is important to reactor design and has been the subject of a large number of experimental and theoretical investigations. The purpose of this work is to review the recent experimental results to obtain a summary of these results and to determine those research areas that require additional information. The investigations have been classified into four subgroups based on the different experimental methods used. These four are: (1) irradiation creep using stress relaxation methods, (2) creep measurements using pressurized tubes, (3) irradiation creep from constant applied load, and (4) irradiation creep experiments using accelerated particles. The similarity and the differences of the results from these methods are discussed and a summary of important results and suggested areas for research is presented. In brief, the important results relate to the dependence of creep on swelling, temperature, stress state and alloying additions. In each of these areas new results have been presented and new questions have arisen which require further research to answer. 65 references

  9. Modelling property changes in graphite irradiated at changing irradiation temperature

    CSIR Research Space (South Africa)

    Kok, S

    2011-01-01

    Full Text Available A new method is proposed to predict the irradiation induced property changes in nuclear; graphite, including the effect of a change in irradiation temperature. The currently used method; to account for changes in irradiation temperature, the scaled...

  10. Study of irradiation creep of vanadium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Strain, R.V.; Smith, D.L. [Argonne National Lab., IL (United States)] [and others

    1997-08-01

    Thin-wall tubing was produced from the 832665 (500 kg) heat of V-4 wt.% Cr-4 wt.% Ti to study its irradiation creep behavior. The specimens, in the form of pressurized capsules, were irradiated in Advanced Test Reactor and High Flux Isotope Reactor experiments (ATR-A1 and HFIR RB-12J, respectively). The ATR-A1 irradiation has been completed and specimens from it will soon be available for postirradiation examination. The RB-12J irradiation is not yet complete.

  11. Verification of thermal-irradiation stress analytical code VIENUS of graphite block

    International Nuclear Information System (INIS)

    Iyoku, Tatsuo; Ishihara, Masahiro; Shiozawa, Shusaku; Shirai, Hiroshi; Minato, Kazuo.

    1992-02-01

    The core graphite components of the High Temperature Engineering Test Reactor (HTTR) show both the dimensional change (irradiation shrinkage) and creep behavior due to fast neutron irradiation under the temperature and the fast neutron irradiation conditions of the HTTR. Therefore, thermal/irradiation stress analytical code, VIENUS, which treats these graphite irradiation behavior, is to be employed in order to design the core components such as fuel block etc. of the HTTR. The VIENUS is a two dimensional finite element viscoelastic stress analytical code to take account of changes in mechanical properties, thermal strain, irradiation-induced dimensional change and creep in the fast neutron irradiation environment. Verification analyses were carried out in order to prove the validity of this code based on the irradiation tests of the 8th OGL-1 fuel assembly and the fuel element of the Peach Bottom reactor. This report describes the outline of the VIENUS code and its verification analyses. (author)

  12. Irradiation creep due to SIPA-induced growth

    International Nuclear Information System (INIS)

    Woo, C.H.

    1980-01-01

    An additional contribution to irradiation creep resulting from the stress-induced preferred adsorption (SIPA) effect is described - SIPA-induced growth (SIG). The mechanism of SIG is discussed and an expression for its contribution to irradiation creep developed. It is shown that SIG is very significant in comparison with SIPA. Enhancement of creep by swelling may also occur. (U.K.)

  13. Changes in creep of polymethylmetacrylate after irradiation

    International Nuclear Information System (INIS)

    Peschanskaya, N.N.; Smolyanskij, A.S.; Suvorova, V.Yu.

    1992-01-01

    A study was made on PMMA, irradiated by different doses of 60 Co γ-radiation in vacuum under creep during compression. It is shown that occurence of tendency to failure at +20 degC is observed at doses of D > 100 kGy (> 10 Mrad), whereas sufficient decrease of deformation before failure takes place at D > 350 kGy. Peculiarities of behaviour of irradiated and nonirradiated PMMA under compression and tension were correlated. It is noted that critical irradiation doses may differ sufficiently for different loading conditions, deformation and longevity characteristics

  14. Irradiation creep in zirconium single crystals

    International Nuclear Information System (INIS)

    MacEwen, S.R.; Fidleris, V.

    1976-07-01

    Two identical single crystals of crystal bar zirconium have been creep tested in reactor. Both specimens were preirradiated at low stress to a dose of about 4 x 10 23 n/m 2 (E > 1 MeV), and were then loaded to 25 MPa. The first specimen was loaded with reactor at full power, the second during a shutdown. The loading strain for both crystals was more than an order of magnitude smaller than that observed when an identical unirradiated crystal was loaded to the same stress. Both crystals exhibited periods of primary creep, after which their creep rates reached nearly constant values when the reactor was at power. During shutdowns the creep rates decreased rapidly with time. Electron microscopy revealed that the irradiation damage consisted of prismatic dislocation loops, approximately 13.5 nm in diameter. Cleared channels, identified as lying on (1010) planes, were also observed. The results are discussed in terms of the current theories for flux enhanced creep in the light of the microstructures observed. (author)

  15. Irradiation creep in ferritic steels

    International Nuclear Information System (INIS)

    Vandermeulen, W.; Bremaecker, A. de; Burbure, S. de; Huet, J.J.; Asbroeck, P. van

    Pressurized and non-pressurized capsules of several ferritic steels have been irradiated in Rapsodie between 400 and 500 0 C up to 3.7 x 10 22 n/cm 2 (E>0.1 MeV). Results of the diameter measurements are presented and show that the total in-pile deformation is lower than for austenitic steels

  16. Irradiation creep, stress relaxation and a mechanical equation of state

    International Nuclear Information System (INIS)

    Foster, J.P.

    1976-01-01

    Irradiation creep and stress relaxation data are available from the United Kingdom for 20 percent CW M316, 20 percent CW FV 548 and FHT PE16 using pure torsion in the absence of swelling at 300 0 C. Irradiation creep models were used to calculate the relaxation and permanent deflection of the stress relaxation tests. Two relationships between irradiation creep and stress relaxation were assessed by comparing the measured and calculated stress relaxation and permanent deflection. The results show that for M316 and FV548, the stress relaxation and deflection may be calculated using irradiation creep models when the stress rate term arising from the irradiation creep model is set equal to zero. In the case of PE16, the inability to calculate the stress relaxation and permanent deflection from the irradiation creep data was attributed to differences in creep behavior arising from lot-to-lot variations in alloying elements and impurity content. A modification of the FV548 and PE16 irradiation creep coefficients was necessary in order to calculate the stress relaxation and deflection. The modifications in FV548 and PE16 irradiation creep properties reduces the large variation in the transient or incubation parameter predicted by irradiation creep tests for M316, FV548 and PE16

  17. Effect of microstructure on light ion irradiation creep in nickel

    International Nuclear Information System (INIS)

    Henager, C.H. Jr.; Simonen, E.P.; Bradley, E.R.; Stang, R.G.

    1983-01-01

    The concept of inhomogeneous slip or localized deformation is introduced to account for a weak dependence of irradiation creep on initial microstructure. Specimens of pure nickel (Ni) with three different microstructures were irradiated at 473 K with 15-17 MeV deuterons in the Pacific Northwest Laboratory (PNL) light ion irradiation creep apparatus. A dispersed barrier model for Climb-Glide (CG) creep was unable to account for the observed creep rates and creep strains. The weak dependence on microstructure was consistent with the Stress Induced Preferential Absorption (SIPA) creep mechanism but a high stress enhanced bias had to be assumed to account for the creep rates. Also, SIPA was unable to account for the observed creep strains. The CG and SIPA modeling utilized rate theory calculations of point defect fluxes and transmission electron microscopy for sink sizes and densities. (orig.)

  18. Effects of microstructure on light ion irradiation creep in nickel

    International Nuclear Information System (INIS)

    Henager, C.H. Jr.; Simonen, E.P.; Bradley, E.R.; Stang, R.G.

    1982-10-01

    The concept of inhomogeneous slip or localized deformation is introduced to account for a weak dependence of irradiation creep on initial microstructure. Specimens of pure Ni with three different microstructures were irradiated at 473 0 K with 15 to 17 MeV deuterons in the PNL light ion irradiation creep apparatus. A dispersed barrier model for climb-glide creep was unable to account for the observed creep rates and creep strains. The weak dependence on microstructure was consistent with the SIPA creep mechanism but a high stress enhanced bias had to be assumed to account for the creep rates. Also, SIPA was unable to account for the observed creep strains. The modeling utilized rate theory calculations of point defect fluxes and transmission electron microscopy for sink sizes and densities

  19. Irradiation-induced amorphization process in graphite

    Energy Technology Data Exchange (ETDEWEB)

    Abe, Hiroaki [Japan Atomic Energy Research Inst., Takasaki, Gunma (Japan). Takasaki Radiation Chemistry Research Establishment

    1996-04-01

    Effects of the element process of irradiation damage on irradiation-induced amorphization processes of graphite was studied. High orientation thermal decomposed graphite was cut about 100 nm width and used as samples. The irradiation experiments are carried out under the conditions of electronic energy of 100-400 KeV, ion energy of 200-600 KeV, ionic species Xe, Ar, Ne, C and He and the irradiation temperature at from room temperature to 900 K. The critical dose ({phi}a) increases exponentially with increasing irradiation temperature. The displacement threshold energy of graphite on c-axis direction was 27 eV and {phi}a{sup e} = 0.5 dpa. dpa is the average number of displacement to atom. The critical dose of ion irradiation ({phi}a{sup i}) was 0.2 dpa at room temperature, and amorphous graphite was produced by less than half of dose of electronic irradiation. Amorphization of graphite depending upon temperature is discussed. (S.Y.)

  20. Irradiation creep under 60 MeV alpha irradiation

    International Nuclear Information System (INIS)

    Reiley, T.C.; Shannon, R.H.; Auble, R.L.

    1980-01-01

    Accelerator-produced charged-particle beams have advantages over neutron irradiation for studying radiation effects in materials, the primary advantage being the ability to control precisely the experimental conditions and improve the accuracy in measuring effects of the irradiation. An apparatus has recently been built at ORNL to exploit this advantage in studying irradiation creep. These experiments employ a beam of 60 MeV alpha particles from the Oak Ridge Isochronous Cyclotron (ORIC). The experimental approach and capabilities of the apparatus are described. The damage cross section, including events associated with inelastic scattering and nuclear reactions, is estimated. The amount of helium that is introduced during the experiments through inelastic processes and through backscattering is reported. Based on the damage rate, the damage processes and the helium-to-dpa ratio, the degree to which fast reactor and fusion reactor conditions may be simulated is discussed. Recent experimental results on the irradiation creep of type 316 stainless steel are presented, and are compared to light ion results obtained elsewhere. These results include the stress and temperature dependence of the formation rate under irradiation. The results are discussed in relation to various irradiation creep mechanisms and to damage microstructure as it evolves during these experiments. (orig.)

  1. Irradiation creep induced anisotropy in a/2 dislocation populations

    International Nuclear Information System (INIS)

    Gelles, D.S.

    1984-05-01

    The contribution of anisotropy in Burgers vector distribution to irradiation creep behavior has been largely ignored in irradiation creep models. However, findings on Frank loops suggest that it may be very important. Procedures are defined to identify the orientations of a/2 Burgers vectors for dislocations in face-centered cubic crystals. By means of these procedures the anisotropy in Burgers vector populations was determined for three Nimonic PE16 pressurized tube specimens irradiated under stress. Considerable anisotropy in Burgers vector population develops during irradiation creep. It is inferred that dislocation motion during irradiation creep is restricted primarily to a climb of a/2 dislocations on 100 planes. Effect of these results on irradiation creep modeling and deformation induced irradiation growth is considered

  2. Theory of void swelling, irradiation creep and growth

    International Nuclear Information System (INIS)

    Wood, M.H.; Bullough, R.; Hayns, M.R.

    Recent progress in our understanding of the fundamental mechanisms involved in swelling, creep and growth of materials subjected to irradiation is reviewed. The topics discussed are: the sink types and their strengths in the lossy continuum; swelling and void distribution analysis, including recent work on void nucleation; and, irradiation creep and growth of zirconium and zircaloy are taken as an example

  3. Characterization of un-irradiated and irradiated reactor graphite; Karakterizacija neozracenog i ozracenog reaktorskog grafita

    Energy Technology Data Exchange (ETDEWEB)

    Marinkovic, S [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1965-11-15

    This report contains three parts: characterization of Yugoslav nuclear graphite development of methods and obtained results, characterization of un-irradiated and irradiated domestic nuclear graphite; calculation of electrical conductivity changes due to vacancies in the graphite crystal lattice.

  4. Creep of fissile ceramic materials under neutron irradiation

    International Nuclear Information System (INIS)

    Brucklacher, D.

    1975-01-01

    Theoretical estimation of the irradiation-induced creep rate of U0 2 by a modification of the Nabarro-Herring model for diffusional creep resulted in a creep rate range between about 6 x 10 -6 to 8 x 10 -5 h -1 for a fission rate of 1 x 10 14 f/cm 3 s and a stress of 2 kgf/mm 2 . Accordingly, the creep rate is enhanced by irradiation at temperatures below 1000 0 to 1200 0 C. It is essentially due to the 'thermal rods' along the fission fragment tracks. Therefore, irradiation-induced creep rates should depend only slightly on temperature and must be markedly lower for carbide and nitride fuel. In-reactor creep experiments on UO 2 were performed at fuel temperatures between 250 0 to 850 0 C. At burnups between 0.3 to 3% the steady-state compressive creep rates are proportional to stress (0 to 4 kgf/mm 2 ) and to fission rate (1 x 10 13 to 2 x 10 14 f/cm 3 s), and are in the range estimated before. The increase in the creep rate with increasing temperature is low and corresponds to an apparent activation energy of only 5200 cal/mol. At burnups above 3 to 4% the stress exponent of the irradiation-induced creep rate increased from n = 1 to n = 1.5. Creep measurements on UO 2 to 15 wt-%Pu0 2 (mechanically mixed, sintered density 86% TD) showed the same temperature dependence as UO 2 below 700 0 C. However, the creep rates were higher by a factor of about 20 compared to fully dense UO 2 . This difference may be explained by assuming a high 'effective' porosity. In-pile creep tests on some UN samples resulted in creep rates that were lower by an order of magnitude than for UO 2 under comparable conditions. (author)

  5. Ion irradiation to simulate neutron irradiation in model graphites: Consequences for nuclear graphite

    Science.gov (United States)

    Galy, N.; Toulhoat, N.; Moncoffre, N.; Pipon, Y.; Bérerd, N.; Ammar, M. R.; Simon, P.; Deldicque, D.; Sainsot, P.

    2017-10-01

    Due to its excellent moderator and reflector qualities, graphite was used in CO2-cooled nuclear reactors such as UNGG (Uranium Naturel-Graphite-Gaz). Neutron irradiation of graphite resulted in the production of 14C which is a key issue radionuclide for the management of the irradiated graphite waste. In order to elucidate the impact of neutron irradiation on 14C behavior, we carried out a systematic investigation of irradiation and its synergistic effects with temperature in Highly Oriented Pyrolitic Graphite (HOPG) model graphite used to simulate the coke grains of nuclear graphite. We used 13C implantation in order to simulate 14C displaced from its original structural site through recoil. The collision of the impinging neutrons with the graphite matrix carbon atoms induces mainly ballistic damage. However, a part of the recoil carbon atom energy is also transferred to the graphite lattice through electronic excitation. The effects of the different irradiation regimes in synergy with temperature were simulated using ion irradiation by varying Sn(nuclear)/Se(electronic) stopping power. Thus, the samples were irradiated with different ions of different energies. The structure modifications were followed by High Resolution Transmission Electron Microscopy (HRTEM) and Raman microspectrometry. The results show that temperature generally counteracts the disordering effects of irradiation but the achieved reordering level strongly depends on the initial structural state of the graphite matrix. Thus, extrapolating to reactor conditions, for an initially highly disordered structure, irradiation at reactor temperatures (200 - 500 °C) should induce almost no change of the initial structure. On the contrary, when the structure is initially less disordered, there should be a "zoning" of the reordering: In "cold" high flux irradiated zones where the ballistic damage is important, the structure should be poorly reordered; In "hot" low flux irradiated zones where the ballistic

  6. Formation of dislocation dipoles in irradiated graphite

    International Nuclear Information System (INIS)

    Niwase, Keisuke

    2005-01-01

    Recently, we have proposed a dislocation dipole accumulation model to explain the irradiation-induced amorphization of graphite. However, the structure of dislocation dipole in the hexagonal networks is still an open question at the atomic-level. In this paper, we propose a possible formation process of the dislocation dipole

  7. CREEP STRAIN CORRELATION FOR IRRADIATED CLADDING

    International Nuclear Information System (INIS)

    P. Macheret

    2001-01-01

    In an attempt to predict the creep deformation of spent nuclear fuel cladding under the repository conditions, different correlations have been developed. One of them, which will be referred to as Murty's correlation in the following, and whose expression is given in Henningson (1998), was developed on the basis of experimental points related to unirradiated Zircaloy cladding (Henningson 1998, p. 56). The objective of this calculation is to adapt Murty's correlation to experimental points pertaining to irradiated Zircaloy cladding. The scope of the calculation is provided by the range of experimental parameters characterized by Zircaloy cladding temperature between 292 C and 420 C, hoop stress between 50 and 630 MPa, and test time extending to 8000 h. As for the burnup of the experimental samples, it ranges between 0.478 and 64 MWd/kgU (i.e., megawatt day per kilogram of uranium), but this is not a parameter of the adapted correlation

  8. Irradiation creep experiments on fusion reactor candidate structural materials

    International Nuclear Information System (INIS)

    Hausen, H.; Cundy, M.R.; Schuele, W.

    1991-01-01

    Irradiation creep rates were determined for annealed and cold-worked AMCR- and 316-type steel alloys in the high flux reactor at Petten, for various irradiation temperatures, stresses and for neutron doses up to 4 dpa. Primary creep elongations were found in all annealed materials. A negative creep elongation was found in cold-worked materials for stresses equal to or below about 100 MPa. An increase of the negative creep elongation is found for decreasing irradiation temperatures and decreasing applied stresses. The stress exponent of the irradiation creep rate in annealed and cold-worked AMCR alloys is n = 1.85 and n = 1.1, respectively. The creep rates of cold-worked AMCR alloys are almost temperature independent over the range investigated (573-693 K). The results obtained in the HFR at Petten are compared with those obtained in ORR and EBR II. The smallest creep rates are found for cold-worked materials of AMCR- and US-PCA-type at Petten which are about a factor two smaller than the creep rates obtained of US-316 at Petten or for US-PCA at ORR or for 316L at EBR II. The scatter band factor for US-PCA, 316L, US-316 irradiated in ORR and EBR II is about 1.5 after a temperature and damage rate normalization

  9. Irradiation creep mechanism: an experimental perspective

    International Nuclear Information System (INIS)

    Garner, F.A.; Gelles, D.S.

    1988-01-01

    The object of this effort is to determine the mechanisms involved in radiation-induced deformation of structural materials and to apply these insights for extrapolation of available fast reactor data to fusion-relevant conditions. An extensive review was conducted of a variety of radiation-induced microstructural data, searching for microstructural records of various irradiation creep mechanisms. It was found that the stress-affected evolution of dislocation microstructure during irradiation is considerably more complex than envisioned in most theoretical modeling studies, particularly in the types of interactive feedback mechanisms operating. Reasonably conclusive evidence was found for a SIPA-type mechanism (stress-induced preferential absorption) operating on both Frank loops and network dislocations. Stress-induced preferential loop nucleation (SIPN) processes may also participate but are thought to be overshadowed by the stronger action of the SIPA-type processes operating on Frank interstitial loops. It was not possible to discern from microstructural evidence between second-order SIPA and first-order SIPA mechanisms, the latter arising from anisotropic diffusion. Evidence was presented, however, that validates the operation of stress-induced preferential unfaulting of Frank loops and stress-induced growth of previously stressed material following removal of applied stress. Dislocation glide mechanisms are also participating but the rate appears to be controlled by SIPA-type climb processes. Applied stresses were shown to generate very anisotropic distributions of Burgers vector in the irradiation-induced microstructure. 108 references, 15 figures, 1 table

  10. Property changes in graphite irradiated at changing irradiation temperature

    International Nuclear Information System (INIS)

    Price, R.J.; Haag, G.

    1979-07-01

    Design data for irradiated graphite are usually presented as families of isothermal curves showing the change in physical property as a function of fast neutron fluence. In this report, procedures for combining isothermal curves to predict behavior under changing irradiation temperatures are compared with experimental data on irradiation-induced changes in dimensions, Young's modulus, thermal conductivity, and thermal expansivity. The suggested procedure fits the data quite well and is physically realistic

  11. Irradiation creep of candidate materials for advanced nuclear plants

    Energy Technology Data Exchange (ETDEWEB)

    Chen, J., E-mail: jiachao.chen@psi.ch; Jung, P.; Hoffelner, W.

    2013-10-15

    In the present paper, irradiation creep results of an intermetallic TiAl alloy and two ferritic oxide dispersion strengthened (ODS) steels are summarized. In situ irradiation creep measurements were performed using homogeneous implantation with α- and p-particles to maximum doses of 0.8 dpa at displacement damage rates of 2–8 × 10{sup −6} dpa/s. The strains of miniaturized flat dog-bone specimens were monitored under uniaxial tensile stresses ranging from 20 to 400 MPa at temperatures of 573, 673 and 773 K, respectively. The effects of material composition, ODS particle size, and bombarding particle on the irradiation creep compliance was studied and results are compared to literature data. Evolution of microstructure during helium implantation was investigated in detail by TEM and is discussed with respect to irradiation creep models.

  12. Analytical and numerical study of graphite IG110 parts in advanced reactor under high temperature and irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Gao, Jinling, E-mail: Jinling_Gao@yeah.net; Yao, Wenjuan, E-mail: wj_yao@yeah.net; Ma, Yudong

    2016-08-15

    Graphical abstract: An analytical model and a numerical procedure are developed to study the mechanical response of IG-110 graphite bricks in HTGR subjected to high temperature and irradiation. The calculation results show great accordance with each other. Rational suggestions on the calculation and design of the IG-110 graphite structure are proposed based on the sensitivity analyses including temperature, irradiation dimensional change, creep and Poisson’s ratio. - Highlights: • Analytical solution of stress and displacement of IG-110 graphite components in HTGR. • Finite element procedure developed for stress analysis of HTGR graphite component. • Parameters analysis of mechanical response of graphite components during the whole life of the reflector. - Abstract: Structural design of nuclear power plant project is an important sub-discipline of civil engineering. Especially after appearance of the fourth generation advanced high temperature gas cooled reactor, structural mechanics in reactor technology becomes a popular subject in structural engineering. As basic ingredients of reflector in reactor, graphite bricks are subjected to high temperature and irradiation and the stress field of graphite structures determines integrity of reflector and makes a great difference to safety of whole structure. In this paper, based on assumptions of elasticity, side reflector is regarded approximately as a straight cylinder structure and primary creep strain is ignored. An analytical study on stress of IG110 graphite parts is present. Meanwhile, a finite element procedure for calculating stresses in the IG110 graphite structure exposed in the high temperature and irradiation is developed. Subsequently, numerical solution of stress in IG110 graphite structure is obtained. Analytical solution agrees well with numerical solution, which indicates that analytical derivation is accurate. Finally, influence of temperature, irradiation dimensional change, creep and Poisson

  13. Ion irradiated graphite exposed to fusion-relevant deuterium plasma

    International Nuclear Information System (INIS)

    Deslandes, Alec; Guenette, Mathew C.; Corr, Cormac S.; Karatchevtseva, Inna; Thomsen, Lars; Ionescu, Mihail; Lumpkin, Gregory R.; Riley, Daniel P.

    2014-01-01

    Graphite samples were irradiated with 5 MeV carbon ions to simulate the damage caused by collision cascades from neutron irradiation in a fusion environment. The ion irradiated graphite samples were then exposed to a deuterium plasma in the linear plasma device, MAGPIE, for a total ion fluence of ∼1 × 10 24 ions m −2 . Raman and near edge X-ray absorption fine structure (NEXAFS) spectroscopy were used to characterize modifications to the graphitic structure. Ion irradiation was observed to decrease the graphitic content and induce disorder in the graphite. Subsequent plasma exposure decreased the graphitic content further. Structural and surface chemistry changes were observed to be greatest for the sample irradiated with the greatest fluence of MeV ions. D retention was measured using elastic recoil detection analysis and showed that ion irradiation increased the amount of retained deuterium in graphite by a factor of four

  14. Irradiation creep and growth of zircaloy-4 tubes

    International Nuclear Information System (INIS)

    Lansiart, S.; Darchis, L.; Pelchat, J.

    1990-01-01

    The influence of temperature and fast neutron flux on irradiation creep and growth of stress relieved zircaloy-4 pressurized tubes has been derived from experimental irradiations in NaK, performed up to 2.5 10 25 n.m -2 in the temperature range [280, 350] 0 C. A significant influence of temperature on axial growth has been observed: at 280 0 C the elongation can no longer be expressed as a linear function of fluence as for the 350 0 C irradiation temperature; diametral growth, on the other hand, always appears negligible. Irradiation creep obviously depends on temperature too; the diametral strain (including thermal part) has been modelled as a sum of primary and secondary terms, the former being independent of fluence. For the tubing considered it is observed that the ranking of the different batches, with respect to diametral creep resistance, is the same before and under irradiation. Concerning axial creep strain the stress relieved material behaves as does an isotropic tube. This is not the case of recrystallized zircaloy-4 F, which shows a non negligible axial deformation, related to the diametral creep one, even though this diametral irradiation creep strain is strongly reduced comparatively to that of the stress relieved material. The comparison of the two materials growth rates is more complex since their dependence on temperature and flux differs

  15. Final report on graphite irradiation test OG-2

    International Nuclear Information System (INIS)

    Price, R.J.; Beavan, L.A.

    1975-01-01

    Results are presented of dimensional, thermal expansivity, thermal conductivity, Young's modulus, and tensile strength measurements on specimens of nuclear graphites irradiated in capsule OG-2. About half the irradiation space was allocated to H-451 near-isotropic petroleum-coke-based graphite or its subsized prototype grade H-429. Most of these specimens had been previously irradiated. Virgin specimens of another near-isotropic graphite, grade TS-1240, were irradiated. Some previously irradiated specimens of needle-coke-based H-327 graphite and pitch-coke-based P 3 JHAN were also included

  16. Irradiation creep transients in Ni-4 at.% Si

    International Nuclear Information System (INIS)

    Nagakawa, J.

    1983-01-01

    In the course of irradiation creep experiments on Ni-4 at.% Si alloy, two types of creep transients were observed on the termination of irradiation. The short term transient was completed within one minute while the long term transient persisted for nearly ten hours. A change in the temperature distribution was excluded from the possible causes, partly because the stress dependence of the observed transient strains was not linear, and partly because the strain increase expected from the temperature change was much smaller than the observed value. Transient behavior of point defects was examined in conjunction with the climb-glide mechanism and the steady-state irradiation creep data. Calculated creep transient due to excess vacancy flux to dislocations was in good agreement with the observed short term transient. The long term transient appears to be a result of dislocation microstructure change. The present results suggest an enhanced irradiation creep under cyclic irradiation conditions which will be encountered in the early generations of fusion reactors. (orig.)

  17. Interaction of irradiation creep and swelling in the creep disappearance regime

    International Nuclear Information System (INIS)

    Garner, F.A.; Toloczko, M.B.

    1992-01-01

    The objective of this effort is to determine the relationship between applied stresses and irradiation-induced dimensional changes in structural metals for fusion applications. Reanalysis of an earlier data set derived from irradiation of long creep tubes in EBR-II at 550 C has shown that the creep-swelling coupling coefficient is relatively independent of temperature at ∼0.6 x 10 -2 MPa -1 , but falls with increases in the swelling rate, especially at high stress levels. The action of stress-affected swelling and carbide precipitation exert different influences on the derivation of this coefficient

  18. Low temperature irradiation creep of tungsten and molybdenum

    International Nuclear Information System (INIS)

    Pouchou, J.-L.

    1975-12-01

    It is demonstrated that the mechanism of stress biased nucleation of dislocation loops may contribute significantly to the low temperature irradiation creep. This is achieved by measuring length and electrical resistivity changes at liquid hydrogen temperature, under fission fragments bombardement. From these measurements (correlated with some electron microscopy observations of irradiated samples), the following three stages of deformation appear: at low doses (smaller than 10 -2 displacement per atom) the deformation is mainly an increase in volume due to point defects. The study of this stage gives the formation volume of a Frenkel pair, and the number of point defects created by an initial fission fragment; for doses higher than 10 -2 d.p.a., the point defects saturation is reached. At this stage, vacancies and interstitials collapse into loops, the nucleation of which is polarized by the applied stress. At zero stress, the corresponding creep rate is zero. At high stresses (>50 kg/mm 2 ), creep rate saturates at value of the order of 10 -21 (FF/cm 3 ) -1 ; because of the recombinations of loops, the creep rate decreases continuously during irradiation. The recombinations lead to a dense dislocation network (formed at doses of 1 d.p.a.), the climb of which oriented by the applied stress gives rise to a steady state creep. The creep rate is smaller, by at least one order of magnitude, than that which is observed in the stage of loop formation [fr

  19. Understanding Creep Mechanisms in Graphite with Experiments, Multiscale Simulations, and Modeling

    International Nuclear Information System (INIS)

    2014-01-01

    Disordering mechanisms in graphite have a long history with conflicting viewpoints. Using Raman and x-ray photon spectroscopy, electron microscopy, x-ray diffraction experiments and atomistic modeling and simulations, the current project has developed a fundamental understanding of early-to-late state radiation damage mechanisms in nuclear reactor grade graphite (NBG-18 and PCEA). We show that the topological defects in graphite play an important role under neutron and ion irradiation.

  20. Understanding Creep Mechanisms in Graphite with Experiments, Multiscale Simulations, and Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Eapen, Jacob [North Carolina State Univ., Raleigh, NC (United States); Murty, Korukonda [North Carolina State Univ., Raleigh, NC (United States); Burchell, Timothy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2014-06-02

    Disordering mechanisms in graphite have a long history with conflicting viewpoints. Using Raman and x-ray photon spectroscopy, electron microscopy, x-ray diffraction experiments and atomistic modeling and simulations, the current project has developed a fundamental understanding of early-to-late state radiation damage mechanisms in nuclear reactor grade graphite (NBG-18 and PCEA). We show that the topological defects in graphite play an important role under neutron and ion irradiation.

  1. Determining the future for irradiated graphite disposal

    International Nuclear Information System (INIS)

    Neighbour, G.B.; Wickham, A.J.; Hacker, P.J.

    2000-01-01

    In recent years, proposals have been made for the long-term treatment of radioactive graphite waste which have ranged from sea dumping through incineration to land-based disposal, sometimes preceded by a variable period of 'safe storage' within the original reactor containment. Nuclear regulators are challenging the proposed length of 'safe storage' on the basis that essential knowledge may be lost. More recently, political constraints have further complicated the issue by eliminating disposal at sea and imposing a 'near-zero release' philosophy, while public opinion is opposed to land-based disposal and has induced a continual drive towards minimizing radioactivity release to the environment from disposal. This paper proposes that, despite various international agreements, it is time to review technically all options for disposal of irradiated graphite waste as a framework for the eventual decision-making process. It is recognized that the socio-economic and political pressures are high and therefore, given that all currently identified options satisfy the present safety limits, the need to minimize the objective risk is shown to be a minor need in comparison to the public's want of demonstrable control, responsiveness and ability to reverse/change the disposal options in the future. Further, it is shown that the eventual decision-making process for a post-dismantling option for graphite waste must optimize the beneficial attributes of subjective risk experienced by the general public. In addition, in advocating and preferred option to the general public, it is recommended that the industry should communicate at a level commensurate with the public understanding and initiate a process of facilitation which enables the public to arrive at their own solution and constituting a social exchange. Otherwise it is concluded that if the indecision over disposal options is allowed to continue then, by default, graphite will remain in long-term supervised storage. (author)

  2. Development of integrated waste management options for irradiated graphite

    Directory of Open Access Journals (Sweden)

    Alan Wareing

    2017-08-01

    Full Text Available The European Treatment and Disposal of Irradiated Graphite and other Carbonaceous Waste project sought to develop best practices in the retrieval, treatment, and disposal of irradiated graphite including other irradiated carbonaceous waste such as structural material made of graphite, nongraphitized carbon bricks, and fuel coatings. Emphasis was given on legacy irradiated graphite, as this represents a significant inventory in respective national waste management programs. This paper provides an overview of the characteristics of graphite irradiated during its use, primarily as a moderator material, within nuclear reactors. It describes the potential techniques applicable to the retrieval, treatment, recycling/reuse, and disposal of these graphite wastes. Considering the lifecycle of nuclear graphite, from manufacture to final disposal, a number of waste management options have been developed. These options consider the techniques and technologies required to address each stage of the lifecycle, such as segregation, treatment, recycle, and ultimate disposal in a radioactive waste repository, providing a toolbox to aid operators and regulators to determine the most appropriate management strategy. It is noted that national waste management programs currently have, or are in the process of developing, respective approaches to irradiated graphite management. The output of the Treatment and Disposal of Irradiated Graphite and other Carbonaceous Waste project is intended to aid these considerations, rather than dictate them.

  3. Development of integrated waste management options for irradiated graphite

    Energy Technology Data Exchange (ETDEWEB)

    Wareing, Alan; Abrahamsen-Mills, Liam; Fowler, Linda; Jarvis, Richard; Banford, Anthony William [National Nuclear Laboratory, Warrington (United Kingdom); Grave, Michael [Doosan Babcock, Gateshead (United Kingdom); Metcalfe, Martin [National Nuclear Laboratory, Gloucestershire (United Kingdom); Norris, Simon [Radioactive Waste Management Limited, Oxon (United Kingdom)

    2017-08-15

    The European Treatment and Disposal of Irradiated Graphite and other Carbonaceous Waste project sought to develop best practices in the retrieval, treatment, and disposal of irradiated graphite including other irradiated carbonaceous waste such as structural material made of graphite, nongraphitized carbon bricks, and fuel coatings. Emphasis was given on legacy irradiated graphite, as this represents a significant inventory in respective national waste management programs. This paper provides an overview of the characteristics of graphite irradiated during its use, primarily as a moderator material, within nuclear reactors. It describes the potential techniques applicable to the retrieval, treatment, recycling/reuse, and disposal of these graphite wastes. Considering the lifecycle of nuclear graphite, from manufacture to final disposal, a number of waste management options have been developed. These options consider the techniques and technologies required to address each stage of the lifecycle, such as segregation, treatment, recycle, and ultimate disposal in a radioactive waste repository, providing a toolbox to aid operators and regulators to determine the most appropriate management strategy. It is noted that national waste management programs currently have, or are in the process of developing, respective approaches to irradiated graphite management. The output of the Treatment and Disposal of Irradiated Graphite and other Carbonaceous Waste project is intended to aid these considerations, rather than dictate them.

  4. Irradiation creep of the mixed oxide UPuO2

    International Nuclear Information System (INIS)

    Combette, Patrick; Milet, Claude

    1976-01-01

    The irradiation creep under compression of the mixed oxide UO 2 -PuO 2 was studied up to fission yields of 6x10 13 fcm -3 s -1 , under stresses -2 , in the temperature range 700-900 deg C. The creep rate is proportional to the applied stress and fission yield, athermal in the studied temperature range and non-dependent of burnup (up to 30000MWjt -1 ). In a sample under compression, swelling is observed due to the formation of fission products during the irradiation and the swelling rate is of the same order that in a cladded fuel element [fr

  5. Removal of 14C from Irradiated Graphite for Graphite Recycle and Waste Volume Reduction

    International Nuclear Information System (INIS)

    Dunzik-Gougar, Mary Lou; Windes, Will; Marsden, Barry

    2014-01-01

    The aim of the research presented here was to identify the chemical form of 14 C in irradiated graphite. A greater understanding of the chemical form of this longest-lived isotope in irradiated graphite will inform not only management of legacy waste, but also development of next generation gas-cooled reactors. Approximately 250,000 metric tons of irradiated graphite waste exists worldwide, with the largest single quantity originating in the Magnox and AGR reactors of UK. The waste quantity is expected to increase with decommissioning of Generation II reactors and deployment of Generation I gas-cooled, graphite moderated reactors. Of greatest concern for long-term disposal of irradiated graphite is carbon-14 14 C, with a half-life of 5730 years.

  6. Actinides in irradiated graphite of RBMK-1500 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Plukienė, R., E-mail: rita@ar.fi.lt; Plukis, A.; Barkauskas, V.; Gudelis, A.; Gvozdaitė, R.; Duškesas, G.; Remeikis, V.

    2014-10-01

    Highlights: • Activation of actinides in the graphite of the RBMK-1500 reactor was analyzed. • Numerical modeling using SCALE 6.1 and MCNPX was used for actinide calculation. • Measurements of the irradiated graphite sample were used for model validation. • Results are important for further decommissioning process of the RBMK type reactors. - Abstract: The activation of graphite in the nuclear power plants is the problem of high importance related with later graphite reprocessing or disposal. The activation of actinide impurities in graphite due to their toxicity determines a particular long term risk to waste management. In this work the activation of actinides in the graphite constructions of the RBMK-1500 reactor is determined by nuclear spectrometry measurements of the irradiated graphite sample from the Ignalina NPP Unit I and by means of numerical modeling using two independent codes SCALE 6.1 (using TRITON-VI sequence) and MCNPX (v2.7 with CINDER). Both models take into account the 3D RBMK-1500 reactor core fragment with explicit graphite construction including a stack and a sleeve but with a different simplification level concerning surrounding graphite and construction of control roads. The verification of the model has been performed by comparing calculated and measured isotope ratios of actinides. Also good prediction capabilities of the actinide activation in the irradiated graphite have been found for both calculation approaches. The initial U impurity concentration in the graphite model has been adjusted taking into account the experimental results. The specific activities of actinides in the irradiated RBMK-1500 graphite constructions have been obtained and differences between numerical simulation results, different structural parts (sleeve and stack) as well as comparison with previous results (Ancius et al., 2005) have been discussed. The obtained results are important for further decommissioning process of the Ignalina NPP and other RBMK

  7. Irradiation creep of 316 and 316 Ti steels

    International Nuclear Information System (INIS)

    Lehmann, J.; Dupouy, J.M.; Boutard, J.L.; Maillard, A.; Broudeur, R.

    1979-07-01

    Irradiation creep results for several 316 and 316 Ti steels show the effects of stress, dose and temperature in the range 400 to 550 0 C. The observed differences are related to the compositioning and metallurgical conditions of the materials. (author)

  8. Irradiation enhanced diffusion and irradiation creep tests in stainless steel alloys

    International Nuclear Information System (INIS)

    Loelgen, R.H.; Cundy, M.R.; Schuele, W.

    1977-01-01

    A review is given of investigations on the rate of phase changes during neutron and electron irradiation in many different fcc alloys showing either precipitation or ordering. The diffusion rate was determined as a function of the irradiation flux, the irradiation temperature and the irradiation dose. It was found that the radiation enhanced diffusion in all the investigated alloys is nearly temperature independent and linearly dependent on the flux. From these results conclusions were drawn concerning the properties of point defects and diffusion mechanisms rate determining during irradiation, which appears to be of a common nature for fcc alloys having a similar structure to those investigated. It has been recognized that the same dependencies which are found for the diffusion rate were also observed for the irradiation creep rate in stainless steels, as reported in literature. On the basis of this obervation a combination of measurements is suggested, of radiation enhanced diffusion and radiation enhanced creep in stainless steel alloys. Measurements of radiation enhanced diffusion are less time consuming and expensive than irradiation creep tests and information on this property can be obtained rather quickly, prior to the selection of stainless steel alloys for creep tests. In order to investigate irradiation creep on many samples at a time two special rigs were developed which are distinguished only by the mode of stress applied to the steel specimens. Finally, a few uniaxial tensile creep tests will be performed in fully instrumented rigs. (Auth.)

  9. Temperature annealing of tracks induced by ion irradiation of graphite

    International Nuclear Information System (INIS)

    Liu, J.; Yao, H.J.; Sun, Y.M.; Duan, J.L.; Hou, M.D.; Mo, D.; Wang, Z.G.; Jin, Y.F.; Abe, H.; Li, Z.C.; Sekimura, N.

    2006-01-01

    Highly oriented pyrolytic graphite (HOPG) samples were irradiated by Xe ions of initial kinetic energy of 3 MeV/u. The irradiations were performed at temperatures of 500 and 800 K. Scanning tunneling microscopy (STM) images show that the tracks occasionally have elongated structures under high-temperature irradiation. The track creation yield at 800 K is by three orders of magnitude smaller compared to that obtained during room-temperature irradiation. STM and Raman spectra show that amorphization occurs in graphite samples irradiated at 500 K to higher fluences, but not at 800 K. The obtained experimental results clearly reveal that the irradiation under high temperature causes track annealing

  10. The behavior of interstitials in irradiated graphite

    International Nuclear Information System (INIS)

    Pedraza, D.F.

    1991-01-01

    A computer model is developed to simulate the behavior of self-interstitials with particular attention to clustering. Owing to the layer structure of graphite, atomistic simulations can be performed using a large parallelepipedic supercell containing a few layers. In particular, interstitial clustering is studied here using a supercell that contains two basal planes only. Frenkel pairs are randomly produced. Interstitials are placed at sites between the crystal planes while vacancies are distributed in the two crystal planes. The size of the computational cell is 20000 atoms and periodic boundary conditions are used in two dimensions. Vacancies are assumed immobile whereas interstitials are given a certain mobility. Two point defect sinks are considered, direct recombination of Frenkel pairs and interstitial clusters. The clusters are assumed to be mobile up to a certain size where they are presumed to become loop nuclei. Clusters can shrink by emission of singly bonded interstitials or by recombination of a peripheral interstitial with a neighboring vacancy. The conditions under which interstitial clustering occurs are reported. It is shown that when clustering occurs the cluster size population gradually shifts towards the largest size cluster. The implications of the present results for irradiation growth and irradiation-induced amorphization are discussed

  11. Irradiation creep by climb-enables glide of dislocations resulting from preferred absorption of point defects

    Energy Technology Data Exchange (ETDEWEB)

    Mansur, L K [Oak Ridge National Lab., TN (USA)

    1979-04-01

    A mechanism of irradiation creep arising from the climb-enabled glide of dislocations due to stress-induced preferred absorption of radiation-produced point defects is proposed. This creep component is here termed preferred absorption glide, PAG. PAG-creep operates in addition to the previously studied components of creep from climb by stress-induced preferred absorption, (SI) PA-creep, and the climb-enabled glide due to excess absorption of interstitials on dislocations during swelling, I-creep. A formulation of the various climb and climb-enabled glide processes which includes earlier results is presented. PAG-creep is comparable in magnitude to PA-creep in the parameter range of applications. While the PSA-creep rate and the I-creep rate are linear in stress, the PAG-creep rate is quadratic in stress and thus dominates at high stresses.

  12. AGC-2 Irradiation Report

    Energy Technology Data Exchange (ETDEWEB)

    Rohrbaugh, David Thomas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Windes, William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Swank, W. David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-06-01

    The Next Generation Nuclear Plant (NGNP) will be a helium-cooled, very high temperature reactor (VHTR) with a large graphite core. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor (HTGR) designs.[ , ] Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphites have been developed and are considered suitable candidates for the new NGNP reactor design. To support the design and licensing of NGNP core components within a commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance are required for the physical, mechanical, and thermal properties of each graphite grade with a specific emphasis on data related to the life limiting effects of irradiation creep on key physical properties of the NGNP candidate graphites. Based on experience with previous graphite core components, the phenomenon of irradiation induced creep within the graphite has been shown to be critical to the total useful lifetime of graphite components. Irradiation induced creep occurs under the simultaneous application of high temperatures, neutron irradiation, and applied stresses within the graphite components. Significant internal stresses within the graphite components can result from a second phenomenon—irradiation induced dimensional change. In this case, the graphite physically changes i.e., first shrinking and then expanding with increasing neutron dose. This disparity in material volume change can induce significant internal stresses within graphite components. Irradiation induced creep relaxes these large internal stresses, thus reducing the risk of crack formation and component failure. Obviously, higher irradiation creep levels tend to relieve more internal stress, thus allowing the

  13. Isotropic nuclear graphites; the effect of neutron irradiation

    International Nuclear Information System (INIS)

    Lore, J.; Buscaillon, A.; Mottet, P.; Micaud, G.

    1977-01-01

    Several isotropic graphites have been manufactured using different forming processes and fillers such as needle coke, regular coke, or pitch coke. Their properties are described in this paper. Specimens of these products have been irradiated in the fast reactor Rapsodie between 400 to 1400 0 C, at fluences up to 1,7.10 21 n.cm -2 PHI.FG. The results show an isotropic behavior under neutron irradiation, but the induced dimensional changes are higher than those of isotropic coke graphites although they are lower than those of conventional extruded graphites made with the same coke

  14. Variation of the properties of siliconized graphite during neutron irradiation

    International Nuclear Information System (INIS)

    Virgil'ev, Y.S.; Chugunova, T.K.; Pikulik, R.G.

    1986-01-01

    The authors evaluate the radiation-induced property changes in siliconized graphite of the industrial grades SG-P and SG-M. The authors simultaneously tested the reference (control) specimens of graphite that are used as the base for obtaining the SG-M siliconized graphite by impregnating with silicon. The suggested scheme (model) atributes the dimensional changes of the siliconized graphite specimens to the effect of the quantitative ratio of the carbide phase and carbon under different conditions of irradiation. If silicon is insufficient for the formation of a dense skeleton, graphite plays a devisive role, and it may be assumed that at an irradiation temperature greater than 600 K, the material shrinks. The presence of isolated carbide inclusions also affects the physicomechanical properties (including the anitfriction properties)

  15. Management of radioactive waste in nuclear power: handling of irradiated graphite from water-cooled graphite reactors

    International Nuclear Information System (INIS)

    Anfimov, S.S.

    2001-01-01

    In this paper an radioactive waste processing of graphite from graphite moderated nuclear reactors at its decommissioning is discussed. Methods of processing of irradiated graphite are presented. It can be concluded that advanced methods for graphite radioactive waste handling are available nowadays. Implementation of these methods will allow to enhance environmental safety of nuclear power that will benefit its progress in the future

  16. Effect of thermal annealing on property changes of neutron-irradiated non-graphitized carbon materials and nuclear graphite

    International Nuclear Information System (INIS)

    Matsuo, Hideto

    1991-06-01

    Changes in dimension of non-graphitized carbon materials and nuclear graphite, and the bulk density, electrical resistivity, Young's modulus and thermal expansivity of nuclear graphite were studied after neutron irradiation at 1128-1483 K and the successive thermal annealing up to 2573 K. Carbon materials showed larger and anisotropic dimensional shrinkage than that of nuclear graphite after the irradiation. The irradiation-induced dimensional shrinkage of carbon materials decreased during annealing at temperatures from 1773 to 2023 K, followed by a slight increase at higher temperatures. On the other hand, the irradiated nuclear graphite hardly showed the changes in length, density and thermal expansivity under the thermal annealing, but the electrical resistivity and Young's modulus showed a gradual decrease with annealing temperature. It has been clarified that there exists significant difference in the effect of thermal annealing on irradiation-induced dimensional shrinkage between graphitized nuclear graphite and non-graphitized carbon materials. (author)

  17. Swelling and irradiation creep of neutron irradiated 316Ti and 15-15Ti steels

    International Nuclear Information System (INIS)

    Maillard, A.; Touron, H.; Seran, J.L.; Chalony, A.

    1992-01-01

    The global behavior, the swelling and irradiation creep resistances of cold worked 316Ti and 15-15Ti, two variants of austenitic steels in use as core component materials of the French fast reactors, are compared. The 15-15Ti leads to a significant improvement due to an increase in the incubation dose swelling. The same phenomena observed on 316Ti are found on 15-15Ti. All species without fuel like samples, wrappers or empty clad swell and creep less than fuel pin cladding irradiated in the same conditions. To explain the swelling difference, as for 316Ti, thermal gradient is also invoked but the irradiation creep difference is not yet clearly understood. To predict the behavior of clads it is indispensable to study the species themselves and to use specific rules. All results confirm the good behavior of 15-15Ti, the best behavior being obtained with the 1% Si doped version irradiated up to 115 dpa

  18. Irradiation induced creep in whiskers of NaCl

    International Nuclear Information System (INIS)

    Khan, J.A.A.

    1977-09-01

    Whiskers of NaCl have been grown and irradiated under flexion by X-rays (approximately 2x10 7 R/h) at room temperature and the residual curvature measured. Complete recovery of the initial form of the whisker within an hour's annealing at 400 0 C proves clearly that the observed deformation (creep) is due to the presence of dislocation loops. The choice of NaCl extremely simplifies the experiment and its interpretation since X-rays create point defects one by one. Moreover, this mode of irradiation, at room temperature, produces a very simple situation: perfect interstitial dislocation loops and immobile point defects which are little influenced by the applied stress. The flexion leads to a stress system which hardly differs from an uniaxial stress. One can study separately the preferential nucleation of dislocation loops and their differential growth by carrying out an irradiation under stress followed by an irradiation without stress and vice versa. It is shown that the induced creep is mostly due to the preferential nucleation of dislocation loops and is little affected by the differential growth of these loops. The nucleation period of the loops is very short: a dose of approximately 10 -5 d.p.a. is largely sufficient for the quasi completion of dislocation loops in a crystal having an impurity concentration of approximately 10 -3 [fr

  19. Spatially resolved nanostructural transformation in graphite under femtosecond laser irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Marcu, A., E-mail: aurelian.marcu@inflpr.ro [National Institute for Laser, Plasma and Radiation Physics, 077125 Bucharest (Romania); Avotina, L. [Institute of Chemical Physics, University of Latvia, Kronvalda 4, LV 1010 Riga (Latvia); Porosnicu, C. [National Institute for Laser, Plasma and Radiation Physics, 077125 Bucharest (Romania); Marin, A. [Ilie Murgulescu” Institute of Physical Chemistry, 202 Splaiul Independentei 060021, Bucharest (Romania); Grigorescu, C.E.A. [National Institute R& D for Optoelectronics INOE 2000, 077125 Bucharest (Romania); Ursescu, D. [National Institute for Laser, Plasma and Radiation Physics, 077125 Bucharest (Romania); Lungu, M. [National Institute of Materials Physics Atomistilor Str., 105 bis, 077125, Magurele (Romania); Demitri, N. [Hard X-ray Beamline and Structural Biology, Elettra-Sincrotrone Trieste, Strada Statale 14 - km 163,5 in AREA Science Park, 34149 Basovizza TS Italy (Italy); Lungu, C.P. [National Institute for Laser, Plasma and Radiation Physics, 077125 Bucharest (Romania)

    2015-11-15

    Graphical abstract: - Highlights: • Polycrystalline graphite was irradiated with a high power fs (IR) laser. • Presence of a diamond peak was detected by synchrotron XRD. • XPS and Raman showed in-depth sp{sup 3}% increase at tens of nm below the surface. • sp{sup 3}% is increasing with laser power density but it is independent of photon absorption rate. • Graphite crystallite size locally increase at tens of nanometers below the irradiated spots. - Abstract: A polycrystalline graphite target was irradiated using infrared (800 nm) femtosecond (120 fs) laser pulses of different energies. Increase of sp{sup 3} bonds percentage and possible diamond crystal formation were investigated ‘in-depth’ and on the irradiated surfaces. Synchrotron X-ray diffraction pattern have shown the presence of a diamond peak in one of the irradiated zones while X-ray photoelectron spectroscopy investigations have shown an increasing tendency of the sp{sup 3} percent in the low power irradiated areas and similarly ‘in the depth’ of the higher power irradiated zones. Multiple wavelength Micro-Raman investigations have confirmed this trend along with an ‘in-depth’ (but not on the surface) increase of the crystallite size. Based on the wavelength dependent photon absorption into graphite, the observed effects are correlated with high density photon per atom and attributed to the melting and recrystallization processes taking place tens of nanometers below the target surface.

  20. Spatially resolved nanostructural transformation in graphite under femtosecond laser irradiation

    International Nuclear Information System (INIS)

    Marcu, A.; Avotina, L.; Porosnicu, C.; Marin, A.; Grigorescu, C.E.A.; Ursescu, D.; Lungu, M.; Demitri, N.; Lungu, C.P.

    2015-01-01

    Graphical abstract: - Highlights: • Polycrystalline graphite was irradiated with a high power fs (IR) laser. • Presence of a diamond peak was detected by synchrotron XRD. • XPS and Raman showed in-depth sp 3 % increase at tens of nm below the surface. • sp 3 % is increasing with laser power density but it is independent of photon absorption rate. • Graphite crystallite size locally increase at tens of nanometers below the irradiated spots. - Abstract: A polycrystalline graphite target was irradiated using infrared (800 nm) femtosecond (120 fs) laser pulses of different energies. Increase of sp 3 bonds percentage and possible diamond crystal formation were investigated ‘in-depth’ and on the irradiated surfaces. Synchrotron X-ray diffraction pattern have shown the presence of a diamond peak in one of the irradiated zones while X-ray photoelectron spectroscopy investigations have shown an increasing tendency of the sp 3 percent in the low power irradiated areas and similarly ‘in the depth’ of the higher power irradiated zones. Multiple wavelength Micro-Raman investigations have confirmed this trend along with an ‘in-depth’ (but not on the surface) increase of the crystallite size. Based on the wavelength dependent photon absorption into graphite, the observed effects are correlated with high density photon per atom and attributed to the melting and recrystallization processes taking place tens of nanometers below the target surface.

  1. Irradiation behavior of graphite shielding materials for FBR

    International Nuclear Information System (INIS)

    Maruyama, Tadashi; Kaito, Takeji; Onose, Shoji; Shibahara, Itaru

    1994-01-01

    Thirteen kinds of isotropic graphites with different density and maximum grain size were irradiated in the experimental fast reactor 'JOYO' to fluences from 2.11 to 2.86x10 26 n/m 2 (E>0.1 MeV) at temperatures from 549 to 597degC. Postirradiation examination was carried out on dimensional change, elastic modulus, and the thermal conductivity. The result of measurement of dimensional change indicated that the graphites irradiated at lower fluences showed shrinkage upon neutron irradiation followed by increase with increasing neutron fluences, irrespective of differences in material parameters. The Young's modulus and Poisson's ratio increased to two to three times of unirradiated values. A large scatter found in Poisson's ratio of unirradiated materials became very small and a linear dependency on density was obtained after irradiation. The thermal conductivity decreased to one-fifth to one-tenth of unirradiated values, but the change in specific heat was negligibly small. The results of postirradiation examination indicated that the changes in physical properties of high density, isotropic graphites were mainly dominated by the irradiation condition rather than their material parameters. Namely, the effects of irradiation induced defects on physical properties of heavily neutron-irradiated graphites are much larger than that of defects associated with as-fabricated specimens. (author)

  2. Neutron irradiations of polycrystalline graphites at 78 K

    International Nuclear Information System (INIS)

    Bochirol, L.; Bonjour, E.; Pluchery, M.

    1961-01-01

    As studies of resistivity restoration after irradiation by electrons have shown that no noticeable healing of created flaws occurs below 80 K, graphite samples are placed in a pool of boiling liquid nitrogen during irradiation and under a pressure slightly greater than normal pressure. Different values are measured: growth rate of a crystalline parameter, stored energy. The influence of irradiation temperature on damages created by a same dose is discussed [fr

  3. Irradiation creep of solution annealed and coldworked 316 stainless steel

    International Nuclear Information System (INIS)

    Boutard, J.L.; Carteret, Y.; Cauvin, R.; Guerin, Y.; Maillard, A.

    1983-04-01

    Because SA and CW 316 stainless steels were used as standard cladding material, a lot of plastic strain data is now avalaible. Most of it is published and analyzed in term of an irradiation creep modulus A defined as the ratio of the equivalent plastic strain to the product of the equivalent stress by the dose. In fact the experimental data and the theoretical analysis of the in-pile deformation mechanisms show a more complicated situation. The purpose of this paper is to reanalyze our results taking into account this situation. This approach is divided in two parts: 1) the high temperature range (T>=450 0 C) where data come from irradiated pins; 2) the low temperature range (T 0 C) where results come from pressurized tubes irradiated in experimental rigs

  4. Data acquisition system for light-ion irradiation creep experiment

    International Nuclear Information System (INIS)

    Hendrick, P.L.; Whitaker, T.J.

    1979-07-01

    Software was developed for a PDP11V/03-based data acquisition system to support the Light-Ion Irradiation Creep Experiment installed at the University of Washington Tandem Van de Graaff Accelerator. The software consists of a real-time data acquisition and storage program, DAC04, written in assembly language. This program provides for the acquisition of up to 30 chennels at 100 Hz, data averaging before storage on disk, alarming, data table display, and automatic disk switching. All analog data are acquired via an analog-to-digital converter subsystem having a resolution of 14 bits, a maximum throughput of 20 kHz, and an overall system accuracy of +-0.01%. These specifications are considered essential for the long-term measurement of irradiation creep strains and temperatures during the light-ion bombardment of irradiation creep specimens. The software package developed also contains a collection of FORTRAN programs designed to monitor a test while in progress. These programs use the foreground/background feature of the RT-11 operating system. The background programs provide a variety of services. The program, GRAFTR, allows transient data (i.e., prior to averaging) to be graphed at the graphics terminal. The program, GRAFAV, allows averaged data to be read from disk and displayed graphically at the terminal. The program, TYPAV, reads averaged data from disk and displays it at the terminal in tabular form. Other programs allow text messages to be written to disk, read from disk, and allow access to DAC04 initialization data. 5 figures, 18 tables

  5. Irradiation enhanced diffusion and irradiation creep tests in stainless steel alloys

    International Nuclear Information System (INIS)

    Loelgen, R.H.; Cundy, M.R.; Schuele, W.

    1977-01-01

    A review is given of investigations on the rate of phase changes during neutron and electron irradiation in many different fcc alloys showing either precipitation or ordering. The diffusion rate was determined as a function of the irradiation flux, the irradiation temperature and the irradiation dose. It was found that the radiation enhanced diffusion in all the investigated alloys is nearly temperature independent and linearly dependent on the flux. From these results conclusions were drawn concerning the properties of point defects and diffusion mechanisms rate determining during irradiation, which appears to be of a common nature for fcc alloys having a similar structure to those investigated. It has been recognized that the same dependencies which are found for the diffusion rate were also observed for the irradiation creep rate in stainless steels, as reported in literature. On the basis of this observation a combination of measurements is suggested, of radiation enhanced diffusion and radiation enhanced creep in stainless steel alloys. The diffusion tests will be performed at the Euratom Joint Research Centre in Ispra, Italy, and the irradiation creep tests will be carried out in the High Flux Reactor /9/ of the Euratom Joint Research Centre in Petten, The Netherlands. In order to investigate irradiation creep on many samples at a time two special rigs were developed which are distinguished only by the mode of stress applied to the steel specimens. In the first type of rig about 50 samples can be tested uniaxially under tension with various combinations of irradiation temperature and stress. The second type of rig holds up to 70 samples which are tested in bending, again with various combinations of irradiation temperature and stress

  6. Irradiation creep and swelling of various austenitic alloys irradiated in PFR and FFTF

    Energy Technology Data Exchange (ETDEWEB)

    Garner, F.A.; Toloczko, M.B. [Pacific Northwest National Lab., Richland, WA (United States)] [and others

    1996-10-01

    In order to use data from surrogate neutron spectra for fusion applications, it is necessary to analyze the impact of environmental differences on property development. This is of particular importance in the study of irradiation creep and its interactions with void swelling, especially with respect to the difficulty of separation of creep strains from various non-creep strains. As part of an on-going creep data rescue and analysis effort, the current study focuses on comparative irradiations conducted on identical gas-pressurized tubes produced and constructed in the United States from austenitic steels (20% CW 316 and 20% CW D9), but irradiated in either the Prototype Fast Reactor (PFR) in the United Kingdom or the Fast Flux Test Facility in the United States. In PFR, Demountable Subassemblies (DMSA) serving as heat pipes were used without active temperature control. In FFTF the specimens were irradiated with active ({+-}{degrees}5C) temperature control. Whereas the FFTF irradiations involved a series of successive side-by-side irradiation, measurement and reinsertion of the same series of tubes, the PFR experiment utilized simultaneous irradiation at two axial positions in the heat pipe to achieve different fluences at different flux levels. The smaller size of the DMSA also necessitated a separation of the tubes at a given flux level into two groups (low-stress and high-stress) at slightly different axial positions, where the flux between the two groups varied {le}10%. Of particular interest in this study was the potential impact of the two types of separation on the derivation of creep coefficients.

  7. Irradiation creep and creep rupture of titanium-modified austenitic stainless steels and their dependence on cold work level

    International Nuclear Information System (INIS)

    Garner, F.A.; Hamilton, M.L.; Eiholzer, C.R.; Toloczko, M.B.; Kumar, A.S.

    1991-11-01

    A titanium-modified austenitic type stainless steel was tested at three cold work levels to determine its creep and creep rupture properties under both thermal aging and neutron irradiation conditions. Both the thermal and irradiation creep behavior exhibit a complex non-monotonic relationship with cold work level that reflects the competition between a number of stress-sensitive and temperature-dependent microstructural processes. Increasing the degree of cold work to 30% from the conventional 20% level was detrimental to its performance, especially for applications above 550 degrees c. The 20% cold work level is preferable to the 10% level, in terms of both in-reactor creep rupture response and initial strength

  8. Final report on graphite irradiation test OG-3

    International Nuclear Information System (INIS)

    Price, R.J.; Beavan, L.A.

    1977-01-01

    The results of dimensional, thermal expansivity, thermal conductivity, Young's modulus, and tensile strength measurements on graphite specimens irradiated in capsule OG-3 are presented. The graphite grades investigated included near-isotropic H-451 (three different preproduction lots), TS-1240, and SO818; needle coke H-327; and European coal tar pitch coke grades P 3 JHA 2 N, P 3 JHAN, and ASI2-500. Data were obtained in the temperature range 823 0 K to 1673 0 K. The peak fast neutron fluence in the experiment was 3 x 10 25 n/m 3 (E greater than 29 fJ)/sub HTGR/; the total accumulated fluence exceeded 9 x 10 25 n/m 2 on some H-451 specimens and 6 x 10 25 n/m 2 on some TS-1240 specimens. Irradiation-induced dimensional changes on H-451 graphite differed slightly from earlier predictions. For an irradiation temperature of about 1225 0 K, axial shrinkage rates at high fluences were somewhat higher than predicted, and the fluence at which radial expansion started (about 9 x 10 25 n/m 2 at 1275 0 K) was lower. TS-1240 graphite underwent smaller dimensional changes than H-451 graphite, while limited data on SO818 and ASI2-500 graphites showed similar behavior to H-451. P 3 JHAN and P 3 JHA 2 N graphites displayed anisotropic behavior with rapid axial shrinkage. Comparison of dimensional changes between specimens from three logs of H-451 and of TS-1240 graphites showed no significant log-to-log variations for H-451, and small but significant log-to-log variations for TS-1240. The thermal expansivity of the near-isotropic graphites irradiated at 865-1045 0 K first increased by 5 percent to 10 percent and then decreased. At higher irradiation temperatures the thermal expansivity decreased by up to 50 percent. Changes in thermal conductivity were consistent with previously established curves. Specimens which were successively irradiated at two different temperatures took on the saturation conductivity for the new temperature

  9. Characteristics of irradiation creep in the first wall of a fusion reactor

    International Nuclear Information System (INIS)

    Coghlan, W.A.; Mansur, L.K.

    1981-01-01

    A number of significant differences in the irradiation environment of a fusion reactor are expected with respect to the fission reactor irradiation environment. These differences are expected to affect the characteristics of irradiation creep in the fusion reactor. Special conditions of importance are identified as the (1) large number of defects produced per pka, (2) high helium production rate, (3) cyclic operation, (4) unique stress histories, and (5) low temperature operations. Existing experimental data from the fission reactor environment is analyzed to shed light on irradiation creep under fusion conditions. Theoretical considerations are used to deduce additional characteristics of irradiation creep in the fusion reactor environment for which no experimental data are available

  10. Examination of the creep behaviour of ceramic fuel elements under neutron irradiation

    International Nuclear Information System (INIS)

    Brucklacher, D.

    1978-01-01

    This paper examines the creeping of UO 2 , UO 2 -PuO 2 and UN under neutron irradiation. It starts with the experimental results about the relation between the thermal creep rate and the load, the temperature, as well as characteristic material values, stoichiometry, grain size and porosity. These correlation are first qualitatively discussed and then compared with the statements of actual quantitative equations. From the models and theories on which these equations are based a modified Nabarro-Heering-equation results for the correlation between the creep rate of ceramic fuels, stress, temperature and the fission rate. In the experimental part of the examination, length-changes of creep samples of UO 2 , (U,Pu)O 2 and UN were measured in specially developed irradiation creep casings in different reactors. The measuring data were corrected and evaluated considering the thermal expansion effects, irregular temperature distribution and swelling effects in such a way that the dependences of the creep rate of UO 2 , UO 2 -PuO 2 and UN under irradiation on stress, temperature, fission rate, burn-up and porosity is obtained. It shows that creeping of fuels under irradiation at high temperatures is equivalent to thermally activated creeping, while at low temperature the creep rate induced by irradiation is much higher than the condition without irradiation. The increment of oxidic nuclear fuels is greater than in UN, the stress dependence on low burn-up is proportional in both cases, and the influence of temperature is quite small. (orig.) [de

  11. Resonant creep enhancement in austenitic stainless steels due to pulsed irradiation at low doses

    International Nuclear Information System (INIS)

    Kishimoto, N.; Amekura, H.; Saito, T.

    1994-01-01

    Steady-state irradiation creep of austenitic stainless steels has been extensively studied as one of the most important design parameters in fusion reactors. The steady-state irradiation creep has been evaluated using in-pile and light-ion experiments. Those creep compliances of various austenitic steels range in the vicinity of ε/Gσ = 10 -6 ∼10 -5 (dpa sm-bullet MPa) -1 , depending on chemical composition etc. The mechanism of steady-state irradiation creep has been elucidated, essentially in terms of stress-induced preferential absorption of point defects into dislocations, and their climb motion. From this standpoint, low doses such as 10 -3 ∼10 -1 dpa would not give rise to any serious creep, and the irradiation creep may not be a critical issue for the low-dose fusion devices including ITER. It is, however, possible that pulsed irradiation causes different creep behaviors from the steady-state one due to dynamic unbalance of interstitials and vacancies. The authors have actually observed anomalous creep enhancement due to pulsed irradiation in austenitic stainless steels. The resonant behavior of creep indicates that pulsed irradiation may cause significant deformation in austenitic steels even at such low doses and slow pulsing rates, especially for the SA-materials. The first-wall materials in plasma operation of ∼10 2 s may suffer from unexpected transient creep, even in the near-term fusion deices, such as ITER. Though this effect might be a transient effect for a relatively short period, it should be taken into account that the pulsed irradiation makes influences on stress relaxation of the fusion components and on the irradiation fatigue. The mechanism and the relevant behaviors of pulse-induced creep will be discussed in terms of a point-defect model based on the resonant interstitial enrichment

  12. Dependence of irradiation creep on temperature and atom displacements in 20% cold worked type 316 stainless steel

    International Nuclear Information System (INIS)

    Gilbert, E.R.

    1976-04-01

    Irradiation creep studies with pressurized tubes of 20 percent cold worked Type 316 stainless steel were conducted in EBR-2. Results showed that as atom displacements are extended above 5 dpa and temperatures are increased above 375 0 C, the irradiation induced creep rate increases with both increasing atom displacements and increasing temperature. The stress exponent for irradiation induced creep remained near unity. Irradiation-induced effective creep strains up to 1.8 percent were observed without specimen failure. 13 figures

  13. Irradiation creep lifetime analysis on first wall structure materials for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Ma, Bing; Peng, Lei, E-mail: penglei@ustc.edu.cn; Zhang, Xiansheng; Shi, Jingyi; Zhan, Jie

    2017-05-15

    Fusion reactor first wall services on the conditions of high surface heat flux and intense neutron irradiation. For China Fusion Engineering Test Reactor (CFETR) with high duty time factor, it is important to analyze the irradiation effect on the creep lifetime of the main candidate structure materials for first wall, i.e. ferritic/martensitic steel, austenite steel and oxide dispersion strengthened steel. The allowable irradiation creep lifetime was evaluated with Larson-Miller Parameter (LMP) model and finite element method. The results show that the allowable irradiation creep lifetime decreases with increasing of surface heat flux, first wall thickness and inlet coolant temperature. For the current CFETR conceptual design, the lifetime is not limited by thermal creep or irradiation creep, which indicated the room for design parameters optimization.

  14. Erosion of pyrolytic graphite and Ti-doped graphite due to high flux irradiation

    International Nuclear Information System (INIS)

    Ohtsuka, Yusuke; Ohashi, Junpei; Ueda, Yoshio; Isobe, Michiro; Nishikawa, Masahiro

    1997-01-01

    The erosion of pyrolytic graphite and titanium doped graphite RG-Ti above 1,780 K was investigated by 5 keV Ar beam irradiation with the flux from 4x10 19 to 1x10 21 m -2 ·s -1 . The total erosion yields were significantly reduced with the flux. This reduction would be attributed to the reduction of RES (radiation enhanced sublimation) yield, which was observed in the case of isotropic graphite with the flux dependence of RES yield of φ -0.26 (φ: flux) obtained in our previous work. The yield of pyrolytic graphite was roughly 30% higher than that of isotropic graphite below the flux of 10 20 m -2 ·s -1 whereas each yield approached to very close value at the highest flux of 1x10 21 m -2 ·s -1 . This result indicated that the effect of graphite structure on the RES yield, which was apparent in the low flux region, would disappear in the high flux region probably due to the disordering of crystal structure. In the case of irradiation to RG-Ti at 1,780 K, the surface undulations evolved with a mean height of about 3 μm at 1.2x10 20 m -2 ·s -1 , while at higher flux of 8.0x10 20 m -2 ·s -1 they were unrecognizable. These phenomena can be explained by the reduction of RES of graphite parts excluding TiC grains. (author)

  15. Study on creep-fatigue life of irradiated austenitic stainless steel

    International Nuclear Information System (INIS)

    Ioka, Ikuo; Miwa, Yukio; Tsuji, Hirokazu; Yonekawa, Minoru; Takada, Fumiki; Hoshiya, Taiji

    2001-01-01

    The low cycle creep-fatigue test with tensile strain hold of the austenitic stainless steel irradiated to 2 dpa was carried out at 823K in vacuum. The applicability of creep-fatigue life prediction methods to the irradiated specimen was examined. The fatigue life on the irradiated specimen without tensile strain hold time was reduced by a factor of 2-5 in comparison with the unirradiated specimen. The decline in fatigue life of the irradiated specimen with tensile strain hold was almost equal to that of the unirradiated specimen. The creep damage of both unirradiated and irradiated specimens was underestimated by the time fraction rule or the ductility exhaustion rule. The creep damage calculated by the time fraction rule or the ductility exhaustion rule increased by the irradiation. The predictions derived from the linear damage rule are unsafe as compared with the experimental fatigue lives. (author)

  16. On the Thermal Conductivity Change of Matrix Graphite Materials after Neutron Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young-Woo; Yeo, Seunghwan; Kim, Eung-Seon; Sah, Injin; Park, Daegyu; Kim, Youngjun; Cho, Moon Sung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    In this work, the variations of the thermal conductivity of the A3-3 matrix graphite after neutron irradiation is discussed as well as of the IG-110 graphite for comparison. Neutron irradiation of the graphite specimens was carried out as a part of the first irradiation test of KAERI's coated particle fuel specimens by use of Hanaro research reactor. This work can be summarized as follows: 1) In the evaluation of the specific heat of the graphite materials, various literature data were used and the variations of the specific heat data of all the graphite specimens are observed well agreed, irrespectively of the difference in specimens (graphite and matrix graphite and irradiated and un-irradiated). 2) This implies that it should be reasonable that for both structural graphite and fuel matrix graphite, and even for the neuron-irradiated graphite, any of these specific heat data set be used in the calculation of the thermal conductivity. 3) For the irradiated A3-3 matrix graphite specimens, the thermal conductivity decreased on both directions. On the radial direction, the tendency of variation upon temperature is similar to that of unirradiated specimen, i.e., decreasing as the temperature increases. 4) In the German irradiation experiments with A3-27 matrix graphite specimens, the thermal conductivity of the un-irradiated specimen shows a decrease and that of irradiated specimen is nearly constant as the temperature increases. 5) The thermal conductivity of the irradiated IG-110 was considerably decreased compared with that of un-irradiated specimens The difference of the thermal conductivity of un-irradiated and irradiated IG-110 graphite specimens is much larger than that of un-irradiated and irradiated A3-3 matrix graphite specimens.

  17. The utilization of a pressurized-graphite/water/oxygen mixture for irradiated graphite incineration

    International Nuclear Information System (INIS)

    Antonini, G.; Perotin, J.P.; Charlot, P.

    1992-01-01

    The authors demonstrate the interest of the utilization of a pressurized-graphite/water/oxygen mixture in the incineration of irradiated graphite. The aqueous phase comes in the form of a three-dimensional system that traps pressurized oxygen, the pulverulent solid being dispersed at the liquid/gas interfaces. These three-phasic formulations give the following advantages: reduction of the apparent viscosity of the mixture in comparison with a solid/liquid mixture at the same solid concentration; reduction of the solid/liquid interactions; self-pulverizability. thus promoting reduction of the flame length utilization of conventional burners; reduction of the flue gas flow rate; complete thermal destruction of graphite. (author)

  18. Thermal and Irradiation Creep Behavior of a Titanium Aluminide in Advanced Nuclear Plant Environments

    Science.gov (United States)

    Magnusson, Per; Chen, Jiachao; Hoffelner, Wolfgang

    2009-12-01

    Titanium aluminides are well-accepted elevated temperature materials. In conventional applications, their poor oxidation resistance limits the maximum operating temperature. Advanced reactors operate in nonoxidizing environments. This could enlarge the applicability of these materials to higher temperatures. The behavior of a cast gamma-alpha-2 TiAl was investigated under thermal and irradiation conditions. Irradiation creep was studied in beam using helium implantation. Dog-bone samples of dimensions 10 × 2 × 0.2 mm3 were investigated in a temperature range of 300 °C to 500 °C under irradiation, and significant creep strains were detected. At temperatures above 500 °C, thermal creep becomes the predominant mechanism. Thermal creep was investigated at temperatures up to 900 °C without irradiation with samples of the same geometry. The results are compared with other materials considered for advanced fission applications. These are a ferritic oxide-dispersion-strengthened material (PM2000) and the nickel-base superalloy IN617. A better thermal creep behavior than IN617 was found in the entire temperature range. Up to 900 °C, the expected 104 hour stress rupture properties exceeded even those of the ODS alloy. The irradiation creep performance of the titanium aluminide was comparable with the ODS steels. For IN617, no irradiation creep experiments were performed due to the expected low irradiation resistance (swelling, helium embrittlement) of nickel-base alloys.

  19. On the analogy between thermally and irradiation induced creep

    International Nuclear Information System (INIS)

    Cozzarelli, F.A.; Huang, S.

    1977-01-01

    Employing an analogy between thermally induced and irradiation induced creep, physical arguments are used first to deduce a one-dimensional constitutive relation for metals under stress in a high temperature and high neutron flux field. This constitutive relation contains modified superposition integrals in which the temperature and flux dependence of the material parameters is included via the use of two reduced time scales; linear elastic, thermal expansion and swelling terms are also included. A systematic development based on thermodynamics, with the stress, temperature increment and defect density increment as independent variables in the Gibbs free energy, is then employed to obtain general three-dimensional memory integrals for strain; the entropy and coupled energy equation are also obtained. Modified superposition integrals similar to those previously obtained by physical argument are then obtained by substituting special functions into the results of the thermodynamic analysis, and the special case of an isotropic stress power law is examined in detail. (Auth.)

  20. Irradiation creep of solution annealed and cold worked 316 stainless steel

    International Nuclear Information System (INIS)

    Boutard, J.L.; Carteret, Y.; Cauvin, R.; Maillard, A.; Guerin, Y.

    1983-01-01

    Irradiation creep strains obtained in-pile on S.A. and C.W. 316 show a linear creep-swelling correlation, the slope of which is rather insensitive to chemical composition and elements in solid solution. The variation of SIPA component resulting only from the evolution of dislocation density and void growth cannot explain such an empirical correlation. The I-creep term has, on the other hand, the right temperature dependence and order of magnitude. (author)

  1. A SIPA-based theory of irradiation creep in the low swelling rate regime

    International Nuclear Information System (INIS)

    Garner, F.A.; Woo, C.H.

    1991-11-01

    A model is presented which describes the major facets of the relationships between irradiation creep, void swelling and applied stress. The increasing degree of anisotropy in distribution of dislocation Burger's vectors with stress level plays a major role in this model. Although bcc metals are known to creep and swell at lower rates than fcc metals, it is predicted that the creep-swelling coupling coefficient is actually larger

  2. Studies on the graphite rupture under irradiation induced strains

    International Nuclear Information System (INIS)

    Jouquet, G.; Berthion, Y.; L'Homme, A.

    1980-01-01

    Following the RMG experiments (failure of graphite by mechanical effect, i.e. under very high temperature gradient) an experimental program called RWG (Failure of Graphite by WIGNER effect) was initiated in 75 at C.E.A. 3 experiments have been already performed in the OSIRIS reactor at Saclay: RWG 01, 02 and 03. A 4th one, RWG04, is scheduled for the end of 79, may be in collaboration with GERMANY. The aim of the RWG experiments is to induce internal stresses in graphite blocks by irradiation at high temperature which would lead or not to their failure so one could bracket, as tightly as possible, the critical value for failure onset in given experimental conditions

  3. Stress state dependence of transient irradiation creep in 20% cold worked 316 stainless steel

    International Nuclear Information System (INIS)

    Foster, J.P.; Gilbert, E.R.

    1998-01-01

    Irradiation creep tests were performed in fast reactors using the stress states of uniaxial tension, biaxial tension, bending and torsion. In order to compare the saturated transient strain irradiation creep component, the test data were converted to equivalent strain and equivalent stress. The saturated transient irradiation creep component was observed to depend on the stress state. The highest value was exhibited by the uniaxial tension stress state, and the lowest by the torsion stress state. The biaxial tension and bending stress state transient component values were intermediate. This behavior appears to be related to the dislocation or microscopic substructure resulting from fabrication processing and the applied stress direction. (orig.)

  4. Inert annealing of irradiated graphite by inductive heating

    International Nuclear Information System (INIS)

    Botzem, W.; Woerner, J.

    2001-01-01

    Fission neutrons change physical properties of graphite being used in nuclear reactors as moderator and as structural material. The understanding of these effects on an atomic model is expressed by dislocations of carbon atoms within the graphite and the thereby stored energy is known as Wigner Energy. The dismantling of the Pile 1 core may necessitate the thermal treatment of the irradiated but otherwise undamaged graphite. This heat treatment - usually called annealing - initiates the release of stored Wigner Energy in a controlled manner. This energy could otherwise give rise to an increase in temperature under certain conditions during transport or preparation for final storage. In order to prevent such an effect it is intended to anneal the major part of Pile 1 graphite before it is packed into boxes suitable for final disposal. Different heating techniques have been assessed. Inductive heating in an inert atmosphere was selected for installation in the Pile 1 Waste Processing Facility built for the treatment and packaging of the dismantled Pile 1 waste. The graphite blocks will be heated up to 250 deg. C in the annealing ovens, which results in the release of significant amount of the stored energy. External heat sources in a final repository will never heat up the storage boxes to such a temperature. (author)

  5. Technique for measuring irradiation creep in polycrystalline SiC fibers

    Energy Technology Data Exchange (ETDEWEB)

    Youngblood, G.E.; Hamilton, M.L.; Jones, R.H. [Pacific Northwest National Lab., Richland, WA (United States)

    1996-10-01

    A bend stress relaxation (BSR) test has been designed to examine irradiation enhanced creep in polycrystalline SiC fibers being considered for fiber reinforcement in SiC/SiC composite. Thermal creep results on Nicalon-CG and Hi-Nicalon were shown to be consistent with previously published data with Hi-Nicalon showing about a 100{degrees}C improvement in creep resistance. Preliminary data was also obtained on Nicalon-S that demonstrated that its creep resistance is greater than that of Hi-Nicalon.

  6. Creep behaviour and microstructure of the ferritic material No. 1-6770 under irradiation

    International Nuclear Information System (INIS)

    Herschbach, K.; Ehrlich, K.; Materna, E.

    Creep behaviour under irradiation of the ferritic steel-DIN-1-6770 is quite different of austenitic steel behaviour, in particular temperature sensitivity is important and response to stress is non linear. The microstructure stays unchanged

  7. Analysis of creep data from MOTA irradiation of 20% cold worked 316 stainless steel

    International Nuclear Information System (INIS)

    Garner, F.A.; Toloczko, M.B.; Puigh, R.J.

    1992-01-01

    The objective of this study is to provide insight into the relationship between void swelling, irradiation creep and applied stress. This insight will be used to develop irradiation creep correlations for fusion applications. Analysis of creep data for 20% cold worked 316 stainless steel irradiated in FFTF/MOTA demonstrates that creep-swelling coupling coefficient is not a strong function of temperature and can be assumed to be -0.6 x 10 -2 MPa -1 in the range 400-600 C. It appears, however, that the creep compliance B o is a moderately strong function of temperature and alloy composition. The latter dependency arises primarily because derived values of B o unavoidably incorporate precipitation-related strains that cannot be easily separated from contributions arising from true creep. It has also been found that at ∼550-600 C there is an upper limit on the total diametral strain rate at 0.33%/dpa. In contrast to the conclusion of an earlier experiment, this limitation does not arise initially from the total disappearance of creep, however. The creep rate first increases with the onset of swelling and then diminishes as the swelling rate increases, disappearing only when the swelling rate reaches its steady state value

  8. Irradiation creep and growth behavior of Zircaloy-4 inner shell of HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Jong-Ha; Cho, Yeong-Garp; Kim, Jong-In [Korea Atomic Energy Research Inst., Daejeon (Korea, Republic of)

    2012-03-15

    The inner shell of the reflector vessel of HANARO was made of Zircaloy-4 rolled plate. Zircaloy-4 rolled plate shows highly anisotropic behavior by fast neutron irradiation. This paper describes the analysis method for the irradiation induced creep and growth of the inner shell of HANARO. The anisotropic irradiation creep behavior was modeled as uniaxial strain-hardening power law modified by Hill's stress potential and the anisotropic irradiation growth was modeled by using volumetric swelling with anisotropic strain rate. In this study, the irradiation induced creep and growth behavior of the inner shell of the HANARO reflector vessel was re-evaluated. The rolling direction, the fast neutron flux, and the boundary conditions were applied with the same conditions as the actual inner shell. Analysis results show that deformation of the inner shell due to irradiation does not raise any problem for the lifetime of HANARO. (author)

  9. The independence of irradiation creep in austenitic alloys of displacement rate and helium to dpa ratio

    Energy Technology Data Exchange (ETDEWEB)

    Garner, F.A.; Toloczko, M.B. [Pacific Northwest National Lab., Richland, WA (United States); Grossbeck, M.L. [Oak Ridge National Lab., TN (United States)

    1997-04-01

    The majority of high fluence data on the void swelling and irradiation creep of austenitic steels were generated at relatively high displacement rates and relatively low helium/dpa levels that are not characteristic of the conditions anticipated in ITER and other anticipated fusion environments. After reanalyzing the available data, this paper shows that irradiation creep is not directly sensitive to either the helium/dpa ratio or the displacement rate, other than through their possible influence on void swelling, since one component of the irradiation creep rate varies with no correlation to the instantaneous swelling rate. Until recently, however, the non-swelling-related creep component was also thought to exhibit its own strong dependence on displacement rate, increasing at lower fluxes. This perception originally arose from the work of Lewthwaite and Mosedale at temperatures in the 270-350{degrees}C range. More recently this perception was thought to extend to higher irradiation temperatures. It now appears, however, that this interpretation is incorrect, and in fact the steady-state value of the non-swelling component of irradiation creep is actually insensitive to displacement rate. The perceived flux dependence appears to arise from a failure to properly interpret the impact of the transient regime of irradiation creep.

  10. Graphites and composites irradiations for gas cooled reactor core structures

    International Nuclear Information System (INIS)

    Van der Laan, J.G.; Vreeling, J.A.; Buckthorpe, D.E.; Reed, J.

    2008-01-01

    Full text of publication follows. Material investigations are undertaken as part of the European Commission 6. Framework Programme for helium-cooled fission reactors under development like HTR, VHTR, GCFR. The work comprises a range of activities, from (pre-)qualification to screening of newly designed materials. The High Flux Reactor at Petten is the main test bed for the irradiation test programmes of the HTRM/M1, RAPHAEL and ExtreMat Integrated Projects. These projects are supported by the European Commission 5. and 6. Framework Programmes. To a large extent they form the European contribution to the Generation-IV International Forum. NRG is also performing a Materials Test Reactor project to support British Energy in preparing extended operation of their Advanced Gas-cooled Reactors (AGR). Irradiations of commercial and developmental graphite grades for HTR core structures are undertaken in the range of 650 to 950 deg C, with a view to get data on physical and mechanical properties that enable engineering design. Various C- and SiC-based composite materials are considered for support structures or specific components like control rods. Irradiation test matrices are chosen to cover commercial materials, and to provide insight on the behaviour of various fibre and matrix types, and the effects of architecture and manufacturing process. The programme is connected with modelling activities to support data trending, and improve understanding of the material behaviour and micro-structural evolution. The irradiation programme involves products from a large variety of industrial and research partners, and there is strong interaction with other high technology areas with extreme environments like space, electronics and fusion. The project on AGR core structures graphite focuses on the effects of high dose neutron irradiation and simultaneous radiolytic oxidation in a range of 350 to 450 deg C. It is aimed to provide data on graphite properties into the parameter space

  11. IAEA international database on irradiated nuclear graphite properties

    International Nuclear Information System (INIS)

    Burchell, T.D.; Clark, R.E.H.; Stephens, J.A.; Eto, M.; Haag, G.; Hacker, P.; Neighbour, G.B.; Janev, R.K.; Wickham, A.J.

    2000-02-01

    This report describes an IAEA database containing data on the properties of irradiated nuclear graphites. Development and implementation of the graphite database followed initial discussions at an IAEA Specialists' Meeting held in September 1995. The design of the database is based upon developments at the University of Bath (United Kingdom), work which the UK Health and Safety Executive initially supported. The database content and data management policies were determined during two IAEA Consultants' Meetings of nuclear reactor graphite specialists held in 1998 and 1999. The graphite data are relevant to the construction and safety case developments required for new and existing HTR nuclear power plants, and to the development of safety cases for continued operation of existing plants. The database design provides a flexible structure for data archiving and retrieval and employs Microsoft Access 97. An instruction manual is provided within this document for new users, including installation instructions for the database on personal computers running Windows 95/NT 4.0 or higher versions. The data management policies and associated responsibilities are contained in the database Working Arrangement which is included as an Appendix to this report. (author)

  12. Transmission electron-microscopic studies of structural changes in polycrystalline graphite after high temperature irradiation

    International Nuclear Information System (INIS)

    Platonov, P.A.; Gurovich, B.A.; Shtrombakh, Ya.I.; Karpukhin, V.I.

    1985-01-01

    Transmission electron-microscopic investigation of polycrystalline graphite before and after irradiation is carried out. The direct use of graphite samples after ion thinning, as an inquiry subject is the basic peculiarity of the work. Main structural components of MPG-6 graphite before and after irradiation are revealed, the structural mechanism of the reactor graphite destruction under irradiation is demonstrated. The mean values of L αm and L cm crystallite dimensions are determined. Radiation defects, occuring in some crystallites after irradiation are revealed by the dark-field electron microscopy method

  13. Structural evaluation of fast reactor core restraint with irradiation creep-swelling opposition effects

    International Nuclear Information System (INIS)

    Kalinowski, J.E.

    1979-01-01

    Irradiation creep and swelling correlations are derived from primary loading in-reactor experiments in which irradiation creep and swelling act in the same direction. When correlation uncertainty bands are applied in core restraint evaluations, significant variability in sub-assembly behavior is predicted. For example, sub-assemblies in the outer core region where neutron flux and duct temperature gradients are significant exhibit bowing responses ranging from a creep dominated outward bow to a swelling dominated inward bow. Furthermore, solutions based on upper bound and lower bound correlation uncertainty combinations are observed to cross-over indicating that such combinations are physically unrealistic in the assessment of creep-swelling opposition effects. In order to obtain realistic upper and lower bound sub-assembly responses, judgement must be applied in the selection of creep-swelling equation uncertainty combinations. Experimental programs have been defined which will provide the needed basic as well as prototypic creep-swelling opposition data for reference and advanced sub-assembly duct alloys. The first of these is an irradiation of cylindrical capsules subjected to a through-wall temperature gradient. This test which is presently underway in the EBR-II reactor will provide the data needed to refine irradiation creep and swelling correlations and their associated uncertainties when applied to core restraint evaluations. Restrained pin and duct bowing experiments in FFTF have also been defined. These will provide the prototypic data necessary to verify irradiated duct bowing methodology. The results of this experimental program are expected to reduce creep and swelling uncertainties and permit better definition of the design window for load plane gaps. (orig.)

  14. Graphite irradiated by swift heavy ions under grazing incidence

    CERN Document Server

    Liu, J; Müller, C; Neumann, R

    2002-01-01

    Highly oriented pyrolytic graphite is irradiated with various heavy projectiles (Ne, Ni, Zn, Xe and U) in the MeV to GeV energy range under different oblique angles of incidence. Using scanning tunneling microscopy, the impact zones are imaged as hillocks protruding from the surface. The diameter of surface-grazing tracks varies between 3 nm (Ne) and 6 nm (U), which is about twice as large as under normal beam incidence. Exclusively for U and Xe projectiles, grazing tracks exhibit long comet-like tails consisting of successive little bumps indicating that the damage along the ion path is discontinuous even for highest electronic stopping powers.

  15. Relationship between swelling and irradiation creep in cold worked PCA stainless steel to 178 DPA at∼400 degrees C

    International Nuclear Information System (INIS)

    Toloczko, M.B.; Garner, F.A.

    1993-01-01

    At 178 dpa and ∼400 degrees C, the irradiation creep behavior of 20% cold-worked PCA has become dominated by the creep disappearance phenomenon. The total diametral deformation rate has reached the limiting value of 0.33%/dpa at the three highest stress levels. The stress-enhancement of swelling tends to camouflage the onset of creep disappearance, however

  16. Stress relaxation analysis and irradiation creep and swelling in pressure tubes

    International Nuclear Information System (INIS)

    Beeston, J.M.; Burr, T.K.

    1979-01-01

    An analysis is presented of slit width test information on two pressure tubes that had been irradiated in test reactors. The analysis showed that differential swelling stresses and thermal stresses undergo relaxation. The mechanism responsible for the stress relaxation at temperatures less than 700 K was irradiation creep. Irradiation creep in thermal test reactor pressure tubes is evidently greater than it would be at equivalent conditions in fast reactors. The residual stresses observed in the slit width tests varied between 30 and 257 MPa and would act to reduce the operating stresses, thus allowing for increased service life of the tubes as compared with no stress relaxation

  17. Bending creep in the direction perpendicular to grain during microwave irradiation

    International Nuclear Information System (INIS)

    Iida, I.

    1989-01-01

    Bending creep tests in the radial direction perpendicular to the grain were carried out on the thirteen different wood species during the microwave irradiation and during the hot-air drying. The course of moisture content of specimen during creep tests were measured at the same time. And then, relationships between the drying rate and the moisture content, or the creep deflection and the moisture content were investigated and disscussed. Results obtained are as follows : 1) The coefficients of drying rate (K 1 ) during microwave irradiation process were from values of 3.40(hr) -1 to 5.65(hr) -1 for different species. With average value of all woods, there were of 4.73(hr) -1 . Therefore, this value show a value of 5.3 times as much as these of hot-air drying. 2) Creep deflection of woods dried by the microwave heating increase remarkably from the start of the microwave irradiation. 3) Ratio ( y 30 /y m ) of creep deflection y m , in region of ∼30% moisture content, to the maximum creep deflection y m were thought the values differ from each wood species, in no relation with the applied stresses and these values have the constant in a wood. Those were estimated about 0.73 for Icho wood and about 0.44 for Buna wood, and moreover it was about 0.6 with average value for all wood species. Consequently, it was recognized that drying rate became remarkably magnitude value during microwave heating. Creep deflection on the 30% moisture content take beyond about half of the total creep deflection. Conseqently, the large creep deformation developed during the high moisture content process, and it constitute a caractaristic frature of microwave heating

  18. Experimental approach and micro-mechanical modeling of the creep behavior of irradiated zirconium alloys

    International Nuclear Information System (INIS)

    Ribis, J.

    2007-12-01

    The fuel rod cladding, strongly affected by microstructural changes due to irradiation such as high density of dislocation loops, is strained by the end-of-life fuel rod internal pressure and the potential release of fission gases and helium during dry storage. Within the temperature range that is expected during dry interim storage, cladding undergoes long term creep under over-pressure. So, in order to have a predictive approach of the behavior of zirconium alloys cladding in dry storage conditions it is essential to take into account: initial dislocation loops, thermal annealing of loops and creep straining due to over pressure. Specific experiments and modelling for irradiated samples have been developed to improve our knowledge in that field. A Zr-1%Nb-O alloy was studied using fine microstructural investigations and mechanical testing. The observations conducted by transmission electron microscopy show that the high density of loops disappears during a heat treatment. The loop size becomes higher and higher while their density falls. The microhardness tests reveal that the fall of loop density leads to the softening of the irradiated material. During a creep test, both temperature and applied stress are responsible of the disappearance of loops. The loops could be swept by the activation of the basal slip system while the prism slip system is inhibited. Once deprived of loops, the creep properties of the irradiated materials are closed to the non irradiated state, a result whose consequence is a sudden acceleration of the creep rate. Finally, a micro-mechanical modeling based on microscopic deformation mechanisms taking into account experimental dislocation loop analyses and creep test, was used for a predictive approach by constructing a deformation mechanism map of the creep behavior of the irradiated material. (author)

  19. Study by internal friction of curing low temperature irradiation defects in graphite

    International Nuclear Information System (INIS)

    Rouby, Dominique.

    1974-01-01

    Micromechanical properties and anelastic effects of neutrons irradiated graphites at 300 and 77 0 K are investigated by internal friction analysis and elasticity modulus variations. Defects created by irradiation are studied and evolution versus dose and annealing is followed [fr

  20. AGC-2 Graphite Preirradiation Data Package

    Energy Technology Data Exchange (ETDEWEB)

    David Swank; Joseph Lord; David Rohrbaugh; William Windes

    2012-10-01

    The NGNP Graphite R&D program is currently establishing the safe operating envelope of graphite core components for a Very High Temperature Reactor (VHTR) design. The program is generating quantitative data necessary for predicting the behavior and operating performance of the new nuclear graphite grades. To determine the in-service behavior of the graphite for pebble bed and prismatic designs, the Advanced Graphite Creep (AGC) experiment is underway. This experiment is examining the properties and behavior of nuclear grade graphite over a large spectrum of temperatures, neutron fluences and compressive loads. Each experiment consists of over 400 graphite specimens that are characterized prior to irradiation and following irradiation. Six experiments are planned with the first, AGC-1, currently being irradiated in the Advanced Test Reactor (ATR) and pre-irradiation characterization of the second, AGC-2, completed. This data package establishes the readiness of 512 specimens for assembly into the AGC-2 capsule.

  1. Irradiation-induced creep in 316 and 304L stainless steels

    International Nuclear Information System (INIS)

    Walters, L.C.; McVay, G.L.; Hudman, G.D.

    1977-01-01

    Recent results are presented from the in-reactor creep experiments that are being conducted by Argonne National Laboratory. The experiments consist of four subassemblies that contain helium-pressurized as well as unstressed capsules of 316 and 304L stainless steels in several metallurgical conditions. Experiments are being irradiated in row 7 of the EBR-II sodium-cooled fast breeder reactor. Three of the subassemblies are being irradiated at temperatures near 400 0 C, and the fourth subassembly is being irradiated at a temperature of 550 0 C. Creep and swelling strains were determined by profilometer measurements on the full length of the capsules after each irradiation cycle. The accumulated neutron dose on the 304L capsules at 385 0 C was 45 dpa; on the 316 capsules at 400 0 C, 40 dpa; and on the 316 capsules at 550 0 C, 25 dpa. It was found that the in-reactor creep rates were linearly dependent on hoop stress, with the exception being capsules of 316 stainless steel that had been given long-term carbide aging treatment and then irradiated at 550 0 C. Those capsules exhibited much higher creep and swelling rates than their unaged counterparts. For the metallurgical conditions where significant swelling was observed (solution-annealed 304L and aged 316 stainless steels), it was found that the in-reactor creep rates were readily fit to a model that related the creep rates to accumulated swelling. Additionally, it was found that the stress-normalized creep rate for 20%-cold-worked 316 stainless steel at a temperature of 550 0 C was 1.6 times that observed at 400 0 C

  2. The irradiation induced creep in fuel compact materials for H.T.R. applications

    International Nuclear Information System (INIS)

    Veringa, H.; Blackstone, R.; Loelgen, R.

    1976-01-01

    Restrained shrinkage experiments up to 3 x 10 21 ncm -2 (DNE) in the temperature range of 600-1,200 0 C on three different dummy coated particle fuel compact materials were performed in the High Flux Reactor at Petten, the Netherlands. The data were evaluated to obtain the steady state irradiation creep coefficient of the compacts. It was found that for the materials investigated, the creep coefficient is temperature-dependent, but no clear relationship to the Young's modulus could be established. Under certain conditions, this irradiation-induced plasticity influences the elastic properties, while also the creep coefficient increases. This effect coincides with the formation and further opening of cracks due to stresses caused by irradiation shrinkage of the matrix material. (orig.) [de

  3. Irradiation creep of various ferritic alloys irradiated at {approximately}400{degrees}C in the PFR and FFTF reactors

    Energy Technology Data Exchange (ETDEWEB)

    Toloczko, M.B.; Garner, F.A. [Pacific Northwest National Lab., Richland, WA (United States); Eiholzer, C.R. [Westinghouse Hanford Company, Richland, WA (United States)

    1997-04-01

    Three ferritic alloys were irradiated in two fast reactors to doses of 50 dpa or more at temperatures near 400{degrees}C. One martensitic alloy, HT9, was irradiated in both the FFTF and PFR reactors. PFR is the Prototype Fast Reactor in Dourneay, Scotland, and FFTF is the Fast Flux Test Facility in Richland, WA. D57 is a developmental alloy that was irradiated in PFR only, and MA957 is a Y{sub 2}O{sub 3} dispersion-hardened ferritic alloy that was irradiated only in FFTF. These alloys exhibited little or no void swelling at {approximately}400{degrees}C. Depending on the alloy starting condition, these steels develop a variety of non-creep strains early in the irradiation that are associated with phase changes. Each of these alloys creeps at a rate that is significantly lower than that of austenitic steels irradiated in the same experiments. The creep compliance for ferritic alloys in general appears to be {approximately}0.5 x 10{sup {minus}6} MPa{sup {minus}1} dpa{sup {minus}1}, independent of both composition and starting state. The addition of Y{sub 2}O{sub 3} as a dispersoid does not appear to change the creep behavior.

  4. Analysis of irradiation creep and the structural integrity of fusion in-vessel components

    International Nuclear Information System (INIS)

    Karditsas, Panayiotis J.

    2000-01-01

    This paper presents a brief review of the irradiation creep mechanism, analyses of the effect on the performance and behaviour of fusion in-vessel components, and discusses procedures for the estimation of in-service time (or lifetime) of components under combined creep-fatigue. The irradiation creep models and proposed theories are examined and analysed to produce a creep law relevant to fusion conditions. The necessary material data, constitutive equations and other parameters needed for estimation of in-service time from the combination of creep and fatigue damage are identified. Wherever possible, design curves are proposed for stress and strain. Time dependent non-linear elastoplastic example calculations are performed, for a typical first wall structure under power plant loading conditions, assuming austenitic and martensitic steel as structural materials, including material irradiation creep. The results of calculations for the stress and strain history of the first wall are used together with the proposed cumulative damage expressions derived in this study to estimate the in-service time, including the effects of stress relaxation due to creep, reduction of ductility (or fracture strain) and helium-to-displacement-damage ratio. The calculations give a displacement damage of ∼70 dpa for the 316 austenitic steel and ∼110-130 dpa for the martensitic steel. Provided there are no power transients, for a design strain of 0.5%, the in-service time is estimated to be ∼3 years for the 316 steel case (at 2.2 MW/m 2 wall load) and the high wall loading martensitic steel (5.0 MW/m 2 case), and ∼5.3 years for the martensitic steel at lower wall load (2.2 MW/m 2 case). The difficulty in defending these results lies in the uncertainty arising from the limited database and experience of the material properties, especially the creep constitutive law, when exposed to fusion environments

  5. Microstructures of beta-silicon carbide after irradiation creep deformation at elevated temperatures

    International Nuclear Information System (INIS)

    Katoh, Yutai; Kondo, Sosuke; Snead, Lance L.

    2008-01-01

    Microstructures of silicon carbide were examined by transmission electron microscopy (TEM) after creep deformation under neutron irradiation. Thin strip specimens of polycrystalline and monocrystalline, chemically vapor-deposited, beta-phase silicon carbide were irradiated in the high flux isotope reactor to 0.7-4.2 dpa at nominal temperatures of 640-1080 deg. C in an elastically pre-strained bend stress relaxation configuration with the initial stress of ∼100 MPa. Irradiation creep caused permanent strains of 0.6 to 2.3 x 10 -4 . Tensile-loaded near-surface portions of the crept specimens were examined by TEM. The main microstructural features observed were dislocation loops in all samples, and appeared similar to those observed in samples irradiated in non-stressed conditions. Slight but statistically significant anisotropy in dislocation loop microstructure was observed in one irradiation condition, and accounted for at least a fraction of the creep strain derived from the stress relaxation. The estimated total volume of loops accounted for 10-45% of the estimated total swelling. The results imply that the early irradiation creep deformation of SiC observed in this work was driven by anisotropic evolutions of extrinsic dislocation loops and matrix defects with undetectable sizes

  6. Changes in the physical and mechanical properties of graphite on irradiation in ditolylmethane

    International Nuclear Information System (INIS)

    Gavrilin, A.I.; Lebedev, I.G.; Sudakova, N.V.; Rizvanov, V.K.

    1987-01-01

    Results are presented from the irradiation and mechanical and structural testing of four grades of graphite - GMZ, VPG, MPG-6, and PG-50 - for use as moderator materials in organic cooled and graphite moderated reactors. Irradiation was carried out in the ARBUS-AST-1 reactor. Photomicrography was used to determine pore structure and ultimate strength in bending and compression was determined mechanically. Irradiation was found to increase the strength of GMZ, PMG-6, and PG-50 considerably, due to the healing of microdefects as a result of the pores filling with radiolysis products from the coolant, ditolylmethane. Conversely, VPG graphite, which has closed porosity, lost strength on irradiation

  7. Effects of prior stress history on the irradiation creep of 20% cold-worked AISI 316 stainless steel

    International Nuclear Information System (INIS)

    Chin, B.A.; Straalsund, J.L.; Wire, G.L.

    1979-01-01

    The following conclusions resulted from this study: An in-reactor transient component of creep is found to occur whenever the stress level is increased. The transient is principally a thermal process, short in duration, and only weakly dependent on flux. The observed irradiation component of in-reactor creep is independent of prior stress history. Microstructural development during irradiation is influenced predominantly by the irradiation flux and temperature variables, and only to a minor extent by the irradiation stress history. (Auth.)

  8. Effects of the temperature and the irradiation on the behaviour of chlorine 37 in nuclear graphite: consequences on the mobility of chlorine 36 in irradiated graphites

    International Nuclear Information System (INIS)

    Blondel, Antoine

    2013-01-01

    This thesis deals with the studies of the management of irradiated graphite wastes issued from the dismantling of the UNGG French reactors. This work focuses on the behavior of 36 Cl. This radionuclide is mainly issued through the neutron activation of 35 Cl by the reaction 35 Cl(n, γ) 36 Cl, pristine chlorine being an impurity of nuclear graphite, present at the level of some at.ppm. 36 Cl is a long lived radionuclide (about 300,000 years) and is highly soluble in water and mobile in concrete and clay. The solubilization of 36 Cl is controlled by the water accessibility into irradiated graphite pores as well as by factors related to 36 Cl itself such as its chemical speciation and its location within the irradiated graphite. Both speciation and chlorine location should strongly influence its behaviour and need to be taken into account for the choice of liable management options. However, data on radioactive chlorine features are difficult to assess in irradiated graphite and are mainly related to detection sensitivity problems. In this context, we simulated and evaluated the impact of the temperature, the irradiation and the radiolytic oxidation on the chlorine 36 behaviour. In order to simulate the presence of 36 Cl, we implanted 37 Cl into virgin nuclear graphite. Ion implantation has been widely used to study the lattice location, the diffusion and the release of fission and activation products in nuclear materials. Our results on the comparative effects of the temperature and the irradiation show that chlorine occurs in irradiated graphite on temperature and electronic and nuclear irradiation improve this effect. (author)

  9. The compositional dependence of irradiation creep of austenitic alloys irradiated in PFR at 420{degrees}C

    Energy Technology Data Exchange (ETDEWEB)

    Toloczko, M.B.; Garner, F.A. [Pacific Northwest National Lab., Richland, WA (United States); Munro, B. [AEA Technology, Dounreay (United Kingdom)] [and others

    1997-04-01

    Irradiation creep data are expensive and often difficult to obtain, especially when compared to swelling data. This requires that maximum use be made of available data sources in order to elucidate the parametric dependencies of irradiation creep for application to new alloys and to new environments such as those of proposed fusion environments. One previously untapped source of creep data is that of a joint U.S./U.K. experiment conducted in the Prototype Fast Reactor (PFR) in Dounreay, Scotland. In this experiment, five austenitic steels were irradiated in a variety of starting conditions. In particular, these steels spanned a large range (15-40%) of nickel contents, and contained strong variations in Mo, Ti, Al, and Nb. Some alloys were solution-strengthened and some were precipitation-strengthened. Several were cold-worked. These previously unanalyzed data show that at 420{degrees}C all austenitic steels have a creep compliance that is roughly independent of the composition of the steel at 2{+-}1 x 10{sup {minus}6}MPa{sup {minus}1} dpa{sup {minus}1}. The variation within this range may arise from the inability to completely separate the non-creep strains arising from precipitation reactions and the stress-enhancement of swelling. Each of these can be very sensitive to the composition and starting treatment of a steel.

  10. Irradiation-induced creep in fuel compacts for high-temperature reactor applications

    Energy Technology Data Exchange (ETDEWEB)

    Veringa, H; Blackstone, R [Stichting Energieonderzoek Centrum Nederland, Petten; Loelgen, R

    1977-01-01

    Restrained shrinkage experiments at neutron fluences up to 3 x 10/sup 21/ n cm/sup -2/ DNE in the temperature range 600 to 1200/sup 0/C were performed on three different dummy coated-particle fuel compacts in the high-flux reactor at Petten. The data were evaluated to obtain the steady-state radiation creep coefficient of the compacts. It was found that, for the materials investigated, the creep coefficient is temperature dependent, but no clear relationship with Young's modulus could be established. Under certain conditions this irradiation-induced plasticity influences the elastic properties, with the concomitant increase of the creep coefficient. This effect coincides with the formation and further opening up of cracks due to stresses caused by irradiation-induced shrinkage of matrix material.

  11. Irradiation-induced creep in fuel compacts for high-temperature reactor applications

    International Nuclear Information System (INIS)

    Veringa, H.; Blackstone, R.; Loelgen, R.

    1977-01-01

    Restrained shrinkage experiments at neutron fluences up to 3 x 10 21 n cm -2 DNE in the temperature range 600 to 1200 0 C were performed on three different dummy coated-particle fuel compacts in the high-flux reactor at Petten. The data were evaluated to obtain the steady-state radiation creep coefficient of the compacts. It was found that, for the materials investigated, the creep coefficient is temperature dependent, but no clear relationship with Young's modulus could be established. Under certain conditions this irradiation-induced plasticity influences the elastic properties, with the concomitant increase of the creep coefficient. This effect coincides with the formation and further opening up of cracks due to stresses caused by irradiation-induced shrinkage of matrix material. (author)

  12. Derivation of a radionuclide inventory for irradiated graphite-chlorine-36 inventory determination

    International Nuclear Information System (INIS)

    Brown, F.J.; Palmer, J.D.; Wood, P.

    2001-01-01

    The irradiation of materials in nuclear reactors results in neutron activation of component elements. Irradiated graphite wastes arise from their use in UK gas-cooled research and commercial reactor cores, and in fuel element components, where the graphite has acted as the neutron moderator. During irradiation the residual chlorine, which was used to purify the graphite during manufacture, is activated to chlorine-36. This isotope is long-lived and poorly retarded by geological barriers, and may therefore be a key radionuclide with respect to post-closure disposal facilities performance. United Kingdom Nirex Limited, currently responsible for the development of a disposal route for intermediate-level radioactive wastes in the UK, carried out a major research programme to support an overall assessment of the chlorine-36 activity of all wastes including graphite reactor components. The various UK gas cooled reactors reactors have used a range of graphite components made from diverse graphite types; this has necessitated a systematic programme to cover the wide range of graphite and production processes. The programme consisted of: precursor measurements - on the surface and/or bulk of representative samples of relevant materials, using specially developed methods; transfer studies - to quantify the potential for transfer of Cl-36 into and between waste streams during irradiation of graphite; theoretical assessments - to support the calculational methodology; actual measurements - to confirm the modelling. For graphite, a total of 458 measurements on samples from 57 batches were performed, to provide a detailed understanding of the composition of nuclear graphite. The work has resulted in the generation of probability density functions (PDF) for the mean chlorine concentration of three classes of graphite: fuel element graphite; Magnox moderator and reflector graphite and AGR reflector graphite; AGR moderator graphite. Transfer studies have shown that a significant

  13. Some physical methods for study of irradiation effects in graphite; Quelques procedes physiques pour etudier les effets de l'irradiation du graphite

    Energy Technology Data Exchange (ETDEWEB)

    Mayer, G; Lecomte, M; Mattmuller, R [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1959-07-01

    A calibration method for a classical apparatus for differential thermal analysis is described in detail. This method achieves a relative precision of 5 per cent in the measurement of the internal energy release accompanying the annealing of irradiated graphites. Elastic constants of graphites are obtained from the frequencies of the longitudinal modes of vibration; procedures for excitation and detection of these vibrations at any temperature between -190 deg. C and +1500 deg. C are described. A procedure for obtaining easily measured deformations of graphites after relatively little irradiation with thermal neutrons is discussed. An application of this method to the study of the thermal annealing of elongation caused by displaced atoms is indicated. (author) [French] On decrit en detail une methode d'etalonnage pour un appareil classique d'analyse thermique differentielle. Cette methode permet de mesurer avec une precision relative de 5% la liberation d'energie interne qui accompagne le 'recuit' des graphites irradies. On deduit les constantes elastiques des graphites des frequences des vibrations longitudinales et on decrit les procedes pour exciter et detecter ces vibrations a toutes les temperatures comprises entre -190 deg. C et + 1500 deg. C. On discute un procede pour obtenir une des deformations de graphites facilement mesurables apres une irradiation relativement faible a l'aide de neutrons thermiques. Une application de cette methode a l'etude du 'recuit' thermique de l'elongation causee par les atomes deplaces est indiquee. (auteur)

  14. Studies on the behavior of graphite structures irradiated in the Dragon Reactor. Dragon Project report

    Energy Technology Data Exchange (ETDEWEB)

    Everett, M. R.; Graham, L. W.; Ridealgh, F.

    1971-11-15

    Design data for the physical and mechanical property changes which occur in graphite structural and fuel body components irradiated in an HTR are largely obtained from small specimens tested in the laboratory and in materials test reactors. A brief data summary is given. This graphite physics data can be used to predict dimensional changes, internal stress generation and strength changes in the graphite materials of HTR fuel elements irradiated in the Dragon Reactor. In this paper, the results which have been obtained from post-irradiation examination of a number of fuel pins, are compared with prediction.

  15. Irradiation-induced creep in metallic nanolaminates characterized by In situ TEM pillar nanocompression

    Energy Technology Data Exchange (ETDEWEB)

    Dillon, Shen J., E-mail: sdillon@illinois.edu [Department of Materials Science and Engineering, University of Illinois Urbana-Champagin, Urbana, IL 61801 (United States); Bufford, Daniel C. [Sandia National Laboratories, Albuquerque, NM 87185 (United States); Jawaharram, Gowtham S.; Liu, Xuying; Lear, Calvin [Department of Materials Science and Engineering, University of Illinois Urbana-Champagin, Urbana, IL 61801 (United States); Hattar, Khalid [Sandia National Laboratories, Albuquerque, NM 87185 (United States); Averback, Robert S. [Department of Materials Science and Engineering, University of Illinois Urbana-Champagin, Urbana, IL 61801 (United States)

    2017-07-15

    This work reports on irradiation-induced creep (IIC) measured on nanolaminate (Cu-W and Ni-Ag) and nanocrystalline alloys (Cu-W) at room temperature using a combination of heavy ion irradiation and nanopillar compression performed concurrently in situ in a transmission electron microscope. Appreciable IIC is observed in multilayers with 50 nm layer thicknesses at high stress, ≈½ the yield strength, but not in multilayers with only 5 nm layer thicknesses.

  16. AGC-3 Graphite Preirradiation Data Analysis Report

    Energy Technology Data Exchange (ETDEWEB)

    William Windes; David Swank; David Rohrbaugh; Joseph Lord

    2013-09-01

    This report describes the specimen loading order and documents all pre-irradiation examination material property measurement data for the graphite specimens contained within the third Advanced Graphite Capsule (AGC-3) irradiation capsule. The AGC-3 capsule is third in six planned irradiation capsules comprising the Advanced Graphite Creep (AGC) test series. The AGC test series is used to irradiate graphite specimens allowing quantitative data necessary for predicting the irradiation behavior and operating performance of new nuclear graphite grades to be generated which will ascertain the in-service behavior of the graphite for pebble bed and prismatic Very High Temperature Reactor (VHTR) designs. The general design of AGC-3 test capsule is similar to the AGC-2 test capsule, material property tests were conducted on graphite specimens prior to loading into the AGC-3 irradiation assembly. However the 6 major nuclear graphite grades in AGC-2 were modified; two previous graphite grades (IG-430 and H-451) were eliminated and one was added (Mersen’s 2114 was added). Specimen testing from three graphite grades (PCEA, 2114, and NBG-17) was conducted at Idaho National Laboratory (INL) and specimen testing for two grades (IG-110 and NBG-18) were conducted at Oak Ridge National Laboratory (ORNL) from May 2011 to July 2013. This report also details the specimen loading methodology for the graphite specimens inside the AGC-3 irradiation capsule. The AGC-3 capsule design requires "matched pair" creep specimens that have similar dose levels above and below the neutron flux profile mid-plane to provide similar specimens with and without an applied load. This document utilized the neutron flux profile calculated for the AGC-3 capsule design, the capsule dimensions, and the size (length) of the selected graphite and silicon carbide samples to create a stacking order that can produce "matched pairs" of graphite samples above and below the AGC-3 capsule elevation mid-point to

  17. Graphite Isotope Ratio Method Development Report: Irradiation Test Demonstration of Uranium as a Low Fluence Indicator

    International Nuclear Information System (INIS)

    Reid, B.D.; Gerlach, D.C.; Love, E.F.; McNeece, J.P.; Livingston, J.V.; Greenwood, L.R.; Petersen, S.L.; Morgan, W.C.

    1999-01-01

    This report describes an irradiation test designed to investigate the suitability of uranium as a graphite isotope ratio method (GIRM) low fluence indicator. GIRM is a demonstrated concept that gives a graphite-moderated reactor's lifetime production based on measuring changes in the isotopic ratio of elements known to exist in trace quantities within reactor-grade graphite. Appendix I of this report provides a tutorial on the GIRM concept

  18. Study on "1"4C content in post-irradiation graphite spheres of HTR-10

    International Nuclear Information System (INIS)

    Wang Shouang; Pi Yue; Xie Feng; Li Hong; Cao Jianzhu

    2014-01-01

    Since the production mechanism of the "1"4C in spherical fuel elements was similar to that of fuel-free graphite spheres, in order to obtain the amount of "1"4C in fuel elements and graphite spheres of HTR-10, the production mechanism of the "1"4C in graphite spheres was studied. The production sources of the "1"4C in graphite spheres and fuel elements were summarized, the amount of "1"4C in the post-irradiation graphite spheres was calculated, the decomposition techniques of graphite spheres were compared, and experimental methods for decomposing the graphite spheres and preparing the "1"4C sample were proposed. The results can lay the foundation for further experimental research and provide theoretical calculations for comparison. (authors)

  19. The study of creep in stainless steel irradiated with fast neutron and alpha particles

    International Nuclear Information System (INIS)

    Correa, D.A.C.

    1985-01-01

    The objective of the present work is to study the creep behavior of the 316 type stainless steel 50% cold worked in different conditions of temperature and applied stress, after neutron radiation and Alfa particles implantation. For this experiment, non-irradiated samples, samples irradiated in the research reactor IEA-R1 with fast neutron (E≥ MeV) up to a fluence of 8.6.10 17 n/cm 2 , and samples implanted with Alfa particles in the cyclotron CV-28 with Helium concentrations of 5 and 26 appm, were creep tested with applied stresses of the 200-300 MPa at temperatures between 650 0 C and 700 0 C. The deformation versus time curves were plotted and it was observed tha the second stage is not well defined, with the creep rate increasing continuously until the occurrence of failure of the material. The study of the effect of increase from 200 MPa to 300 MPa at the same temperature was performed. It can be concluded that this increase produces an approximately 70% reductions in the fracture time of the material, with practically no influence in the total deformation. Samples were tested at different temperatures (650, 675 and 700 0 C) at a same applied stress (200 MPa). It has been observed that a temperature of 50 0 C produces 98,9% of reduction in the fracture time and almost doubles the total deformation. On neutron irradiated samples, creep tests were performed at the same temperature and stress of the non irradiated samples. Comparing the results obtained a tendency of embrittlement due to the neutron irradiation can be observed; no remarkable structure changes were detected due to small fast neutron. Microstructural and metalographic observations were performed before and after each creep test. (author) [pt

  20. Influence of irradiation on high-strength graphites

    International Nuclear Information System (INIS)

    Virgil'ev, Yu.S.; Grebennik, V.N.; Kalyagina, I.P.

    1989-01-01

    To ensure efficiency of the graphite elements of the construction of the masonry of reactors, the graphite must possess high radiation stability, strength, and heat resistance. In this connection, the physical properties of graphites based on uncalcined petroleum coke with a binder - high-temperature hard coal pitch - the amount of which reaches 40% are considered in this paper

  1. The reaction of unirradiated and irradiated nuclear graphites with water vapor in helium

    International Nuclear Information System (INIS)

    Imai, Hisashi; Nomura, Shinzo; Kurosawa, Takeshi; Fujii, Kimio; Sasaki, Yasuichi

    1980-10-01

    Nuclear graphites more than 10 brands were oxidized with water vapor in helium and then some selected graphites were irradiated with fast neutron in the Japan Materials Testing Reactor to clarify the effect of radiation damage of graphite on their reaction behaviors. The reaction was carried out under a well defined condition in the temperature range 800 -- 1000 0 C at concentrations of water vapor 0.38 -- 1.30 volume percent in helium flow of total pressure of 1 atm. The chemical reactivity of graphite irradiated at 1000 +- 50 0 C increased linearly with neutron fluence until irradiation of 3.2 x 10 21 n/cm 2 . The activation energy for the reaction was found to decrease with neutron fluence for almost all the graphites, except for a few ones. The order of reaction increased from 0.5 for the unirradiated graphite to 1.0 for the graphite irradiated up to 6.0 x 10 20 n/cm 2 . Experiment was also performed to study a superposed effect between the influence of radiation damage of graphite and the catalytic action of barium on the reaction rate, as well as the effect of catalyser of barium. It was shown that these effects were not superposed upon each other, although barium had a strong catalytic action on the reaction. (author)

  2. Production of nanodiamonds by high-energy ion irradiation of graphite at room temperature

    International Nuclear Information System (INIS)

    Daulton, T.L.; Kirk, M.A.; Lewis, R.S.; Rehn, L.E.

    2001-01-01

    It has previously been shown that graphite can be transformed into diamond by MeV electron and ion irradiation at temperatures above approximately 600 deg. C. However, there exists geological evidence suggesting that carbonaceous materials can be transformed to diamond by irradiation at substantially lower temperatures. For example, submicron-size diamond aggregates have been found in uranium-rich, Precambrian carbonaceous deposits that never experienced high temperature or pressure. To test if diamonds can be formed at lower irradiation temperatures, sheets of fine-grain polycrystalline graphite were bombarded at 20 deg. C with 350±50 MeV Kr ions to fluences of 6x10 12 cm -2 using the Argonne tandem linear accelerator system (ATLAS). Ion-irradiated (and unirradiated control) graphite specimens were then subjected to acid dissolution treatments to remove untransformed graphite and isolate diamonds that were produced; these acid residues were subsequently characterized by high-resolution and analytical electron microscopy. The acid residue of the ion-irradiated graphite was found to contain nanodiamonds, demonstrating that ion irradiation of graphite at ambient temperature can produce diamond. The diamond yield under our irradiation conditions is low, ∼0.01 diamonds/ion. An important observation that emerges from comparing the present result with previous observations of diamond formation during irradiation is that nanodiamonds form under a surprisingly wide range of irradiation conditions. This propensity may be related to the very small difference in the graphite and diamond free-energies coupled with surface-energy considerations that may alter the relative stability of diamond and graphite at nanometer sizes

  3. Effects of point defect trapping and solute segregation on irradiation-induced swelling and creep

    International Nuclear Information System (INIS)

    Mansur, L.K.

    1978-01-01

    The theory of irradiation swelling and creep, generalized to include impurity trapping of point defects and impurity-induced changes in sink efficiencies for point defects, is reviewed. The mathematical framework is developed and significant results are described. These include the relation between vacancy and interstitial trapping and the effectiveness of trapping as compared to segregation-induced changes in sink efficiencies in modifying void nucleation, void growth, and creep. Current understanding is critically assessed. Several areas requiring further development are identified. In particular those given special attention are the treatment of nondilute solutions and the consequences of current uncertainties in fundamental materials properties whose importance has been identified using the theory

  4. Design and fabrication of a dead weight equipment to perform creep measurements on highly irradiated beryllium specimens

    International Nuclear Information System (INIS)

    Scibetta, M.; Pellettieri, A.; Wouters, P.; Leenaerts, A.; Verpoucke, G.

    2005-01-01

    Beryllium is an important material to be used in the blanket of fusion reactors. It acts as a neutron multiplier that allows tritium production. In order to use this material effectively, some data on creep and swelling behaviour are needed. This paper describes preliminary microstructural investigations and the qualification of a creep set-up that will be used to measure creep of highly irradiated beryllium from the BR2 research reactor matrix. (Author)

  5. IRRADIATION CREEP AND MECHANICAL PROPERTIES OF TWO FERRITIC-MARTENSITIC STEELS IRRADIATED IN THE BN-350 FAST REACTOR

    International Nuclear Information System (INIS)

    Porollo, S. I.; Konobeev, Yu V.; Dvoriashin, A. M.; Budylkin, N. I.; Mironova, E. G.; Leontyeva-Smirnova, M. V.; Loltukhovsky, A. G.; Bochvar, A. A.; Garner, Francis A.

    2002-01-01

    Russian ferritic/martensitic steels EP-450 and EP-823 were irradiated to 20-60 dpa in the BN-350 fast reactor in the form of pressurized creep tubes and small rings used for mechanical property tests. Data derived from these steels serves to enhance our understanding of the general behavior of this class of steels. It appears that these steels exhibit behavior that is very consistent with that of Western steels. Swelling is relatively low at high neutron exposure and confined to temperatures less then 420 degrees C, but may be camouflaged somewhat by precipitation-related densification. The irradiation creep studies confirm that the creep compliance of F/M steels is about one-half that of austenitic steels, and that the loss of strength at test temperatures above 500 degrees C is a problem generic to all F/M steels. This conclusion is supported by post-irradiation measurement of short-term mechanical properties. At temperatures below 500 degrees C both steels retain their high strength (yield stress 0.2=550-600 MPa), but at higher test temperatures a sharp decrease of strength properties occurs. However, the irradiated steels still retain high post-irradiation ductility at test temperatures in the range of 20-700 degrees C.

  6. Graphite

    Science.gov (United States)

    Robinson, Gilpin R.; Hammarstrom, Jane M.; Olson, Donald W.; Schulz, Klaus J.; DeYoung,, John H.; Seal, Robert R.; Bradley, Dwight C.

    2017-12-19

    Graphite is a form of pure carbon that normally occurs as black crystal flakes and masses. It has important properties, such as chemical inertness, thermal stability, high electrical conductivity, and lubricity (slipperiness) that make it suitable for many industrial applications, including electronics, lubricants, metallurgy, and steelmaking. For some of these uses, no suitable substitutes are available. Steelmaking and refractory applications in metallurgy use the largest amount of produced graphite; however, emerging technology uses in large-scale fuel cell, battery, and lightweight high-strength composite applications could substantially increase world demand for graphite.Graphite ores are classified as “amorphous” (microcrystalline), and “crystalline” (“flake” or “lump or chip”) based on the ore’s crystallinity, grain-size, and morphology. All graphite deposits mined today formed from metamorphism of carbonaceous sedimentary rocks, and the ore type is determined by the geologic setting. Thermally metamorphosed coal is the usual source of amorphous graphite. Disseminated crystalline flake graphite is mined from carbonaceous metamorphic rocks, and lump or chip graphite is mined from veins in high-grade metamorphic regions. Because graphite is chemically inert and nontoxic, the main environmental concerns associated with graphite mining are inhalation of fine-grained dusts, including silicate and sulfide mineral particles, and hydrocarbon vapors produced during the mining and processing of ore. Synthetic graphite is manufactured from hydrocarbon sources using high-temperature heat treatment, and it is more expensive to produce than natural graphite.Production of natural graphite is dominated by China, India, and Brazil, which export graphite worldwide. China provides approximately 67 percent of worldwide output of natural graphite, and, as the dominant exporter, has the ability to set world prices. China has significant graphite reserves, and

  7. The influence of low dose irradiation on the creep properties of type 316 welds

    International Nuclear Information System (INIS)

    Marshall, P.; Steeds, J.W.; Lin, Y.P.; Finlan, G.T.

    1987-01-01

    Fully instrumented creep and stress rupture tests have been performed at 873K for times up to 20,000h on a series of type 316 steel/17Cr 8Ni 2Mo weld metal specimens in the unirradiated and thermal neutron irradiated conditions. The specimens tested included all weld metal longitudinal and transverse composites in the as-welded condition and following a stress relief heat treatment of 10h at 1075K. Simulated heat affected zone (HAZ) specimens were also tested. Analysis of the creep results combined with metallography, autoradiography and TEM established that the decrease in properties of irradiated samples is caused by an increasing secondary strain rate due to enhanced helium induced grain boundary fracture of the simulated HAZ and enhanced interdendritic fracture in the weld metal. Implications of strength reductions on the design of welded structures subjected to thermal irradiation are briefly assessed. (author)

  8. Nuclear graphite based on coal tar pitch; behavior under neutron irradiation between 400 and 14000C

    International Nuclear Information System (INIS)

    Mottet, P.; Fillatre, A.; Schill, R.; Micaud, G.

    1977-01-01

    Two nuclear grades of coal tar pitch coke graphites have been developed and tested under neutron irradiation. The neutron irradiation induced dimensional changes between 400 and 1400 0 C, at fluences up to 1,2.10 22 n.cm -2 PHI.FG show a behavior comparable to anisotropic petroleum coke graphites. Less than 10% variation in thermal expansion, maximum decrease by a factor four in thermal conductivity, and large increase of the Young modulus have been observed

  9. Proton irradiated graphite grades for a long baseline neutrino facility experiment

    International Nuclear Information System (INIS)

    Simos, N.; Nocera, P.; Zwaska, R.; Mokhov, N.

    2017-01-01

    In search of a low-Z pion production target for the Long Baseline Neutrino Facility (LBNF) of the Deep Underground Neutrino Experiment (DUNE) four graphite grades were irradiated with protons in the energy range of 140–180 MeV, to peak fluence of ~6.1×10"2"0 p/cm"2 and irradiation temperatures between 120–200 °C. The test array included POCO ZXF-5Q, Toyo-Tanso IG 430, Carbone-Lorraine 2020 and SGL R7650 grades of graphite. Irradiation was performed at the Brookhaven Linear Isotope Producer. Postirradiation analyses were performed with the objective of (a) comparing their response under the postulated irradiation conditions to guide a graphite grade selection for use as a pion target and (b) understanding changes in physical and mechanical properties as well as microstructure that occurred as a result of the achieved fluence and in particular at this low-temperature regime where pion graphite targets are expected to operate. A further goal of the postirradiation evaluation was to establish a proton-neutron correlation damage on graphite that will allow for the use of a wealth of available neutron-based damage data in proton-based studies and applications. Macroscopic postirradiation analyses as well as energy dispersive x-ray diffraction of 200 KeV x rays at the NSLS synchrotron of Brookhaven National Laboratory were employed. The macroscopic analyses revealed differences in the physical and strength properties of the four grades with behavior however under proton irradiation that qualitatively agrees with that reported for graphite under neutrons for the same low temperature regime and in particular the increase of thermal expansion, strength and Young’s modulus. The proton fluence level of ~10"2"0 cm"-"2 where strength reaches a maximum before it begins to decrease at higher fluences has been identified and it agrees with neutron-induced changes. X-ray diffraction analyses of the proton irradiated graphite revealed for the first time the similarity in

  10. Proton irradiated graphite grades for a long baseline neutrino facility experiment

    Science.gov (United States)

    Simos, N.; Nocera, P.; Zhong, Z.; Zwaska, R.; Mokhov, N.; Misek, J.; Ammigan, K.; Hurh, P.; Kotsina, Z.

    2017-07-01

    In search of a low-Z pion production target for the Long Baseline Neutrino Facility (LBNF) of the Deep Underground Neutrino Experiment (DUNE) four graphite grades were irradiated with protons in the energy range of 140-180 MeV, to peak fluence of ˜6.1 ×1020 p /cm2 and irradiation temperatures between 120 - 200 °C . The test array included POCO ZXF-5Q, Toyo-Tanso IG 430, Carbone-Lorraine 2020 and SGL R7650 grades of graphite. Irradiation was performed at the Brookhaven Linear Isotope Producer. Postirradiation analyses were performed with the objective of (a) comparing their response under the postulated irradiation conditions to guide a graphite grade selection for use as a pion target and (b) understanding changes in physical and mechanical properties as well as microstructure that occurred as a result of the achieved fluence and in particular at this low-temperature regime where pion graphite targets are expected to operate. A further goal of the postirradiation evaluation was to establish a proton-neutron correlation damage on graphite that will allow for the use of a wealth of available neutron-based damage data in proton-based studies and applications. Macroscopic postirradiation analyses as well as energy dispersive x-ray diffraction of 200 KeV x rays at the NSLS synchrotron of Brookhaven National Laboratory were employed. The macroscopic analyses revealed differences in the physical and strength properties of the four grades with behavior however under proton irradiation that qualitatively agrees with that reported for graphite under neutrons for the same low temperature regime and in particular the increase of thermal expansion, strength and Young's modulus. The proton fluence level of ˜1020 cm-2 where strength reaches a maximum before it begins to decrease at higher fluences has been identified and it agrees with neutron-induced changes. X-ray diffraction analyses of the proton irradiated graphite revealed for the first time the similarity in

  11. Management of radioactive waste in nuclear power: handling of irradiated graphite from water-cooled graphite reactors

    International Nuclear Information System (INIS)

    Anfimov, S.S.

    2000-01-01

    As a result of decommissioning of water-cooled graphite-moderated reactors, a large amount of rad-waste in the form of graphite stack fragments is generated (on average 1500-2000 tons per reactor). That is why it is essentially important, although complex from the technical point of view, to develop advanced technologies based on up-to-date remotely-controlled systems for unmanned dismantling of the graphite stack containing highly-active long-lived radionuclides and for conditioning of irradiated graphite (IG) for the purposes of transportation and subsequent long term and ecologically safe storage either on NPP sites or in special-purpose geological repositories. The main characteristics critical for radiation and nuclear hazards of the graphite stack are as follows: the graphite stack is contaminated with nuclear fuel that has gotten there as a result of the accidents; the graphite mass is 992 tons, total activity -6?104 Ci (at the time of unit shutdown); the fuel mass in the reactor stack amounts to 100-140 kg, as estimated by IPPE and RDIPE, respectively; γ-radiation dose rate in the stack cells varies from 4 to 4300 R/h, with the prevailing values being in the range from 50 to 100 R/h. In this paper the traditional methods of rad-waste handling as bituminization technology, cementing technology are discussed. In terms of IG handling technology two lines were identified: long-term storage of conditioned IG and IG disposal by means of incineration. The specific cost of graphite immobilization in a radiation-resistant polymeric matrix amounts to -2600 USD per 1 t of graphite, whereas the specific cost of immobilization in slag-stone containers with an inorganic binder (cement) is -1400 USD per 1 t of graphite. On the other hand, volume of conditioned IG rad-waste subject for disposal, if obtained by means of the first technology, is 2-2.5 times less than the volume of rad-waste generated by means of the second technology. It can be concluded from the above that

  12. A theoretical model of accelerated irradiation creep at low temperatures by transient interstitial absorption

    International Nuclear Information System (INIS)

    Stoller, R.E.; Grossbeck, M.L.; Mansur, L.K.

    1990-01-01

    A theoretical model has been developed using the reaction rate theory of radiation effects to explain experimental results that showed higher than expected values of irradiation creep at low temperatures in the Oak Ridge Research Reactor. The customary assumption that the point defect concentrations are at steady state was not made; rather, the time dependence of the vacancy and interstitial concentrations and the creep rate were explicitly calculated. For temperatures below about 100 to 200 degree C, the time required for the vacancy concentration to reach steady state exceeds the duration of the experiment. For example, if materials parameters typical of austenitic stainless steel are used, the calculated vacancy transient dose at 100 degree C is about 100 dpa. At 550 degree C this transient is over by 10 -8 dpa. During the time that the vacancy population remains lower than its steady state value, dislocation climb is increased since defects of primarily one type are being absorbed. Using the time-dependent point defect concentrations, the dislocation climb velocity has been calculated as a function of time and a climb-enabled glide creep model had been invoked. The extended transient time for the vacancies leads to high creep rates at low temperatures. In agreement with the experimental observations, a minimum in the temperature dependence of creep is predicted at a temperature between 50 and 350 degree C. The temperature at which the minimum occurs decreases as the irradiation dose increases. Predicted values of creep at 8 dpa are in good agreement with the results of the ORR-MFE-6J/7J experiment

  13. Determination of Cl-36 in Irradiated Reactor Graphite

    International Nuclear Information System (INIS)

    Beer, H.-F.; Schumann, D.; Stowasser, T.; Hartmann, E.; Kramer, A.

    2016-01-01

    Two of the three research reactors at the Paul Scherrer Institute (PSI), the reactors DIORIT and PROTEUS, contained reactor graphite. Whereas the former research reactor DIORIT has been dismantled completely the PROTEUS is subject to a future decommissioning. In case of the DIORIT the reactor graphite was conditioned applying a procedure developed at PSI. In this case the 36 Cl content had to be determined after the conditioning. The result is reported in this paper. The radionuclide inventory including 36 Cl of the graphite used in PROTEUS was measured and the results are reported in here. It has been proven that the graphite from PROTEUS has a radionuclide inventory near the detection limits. All determined radionuclide activities are far below the Swiss exemptions limits. The graphite from PROTEUS therefore poses no radioactive waste. In contrast, the 36 Cl content of graphite from DIORIT is well above the exemption limits. (author)

  14. Modification of graphite structure by irradiation, revealed by thermal oxidation. Examination by electronic microscopy

    International Nuclear Information System (INIS)

    Rouaud, Michel

    1969-01-01

    Based on the analysis of images obtained by electronic microscopy, this document reports the comparative study of the action of neutrons on three different graphites: a natural one (Ticonderoga) and two pyrolytic ones (Carbone-Lorraine and Raytheon). The approach is based on the modification of features of thermal oxidation of graphites by dry air after irradiation. Different corrosion features are identified. The author states that there seems to be a relationship between the number and shape of these features, and defects existing on the irradiated graphite before oxidation. For low doses, the feature aspect varies with depth at which oxidation occurs. For higher doses, the aspect remains the same [fr

  15. A reverse method for the determination of the radiological inventory of irradiated graphite at reactor scale

    Energy Technology Data Exchange (ETDEWEB)

    Nicaise, Gregory [Institut de Radioprotection et de Surete Nucleaire, Fontenay-aux-roses (France); Poncet, Bernard [EDF-DP2D, Lyon (France)

    2016-11-15

    Irradiated graphite waste will be produced from the decommissioning of the six gas-cooled nuclear reactors operated by Electricite De France (EDF). Determining the radionuclide content of this waste is an important legal commitment for both safety reasons and in order to determine the best suited management strategy. As evidenced by numerous studies nuclear graphite is a very pure material, however, it cannot be considered from an analytical viewpoint as a usual homogeneous material. Because of graphite high purity, radionuclide measurements in irradiated graphite exhibit very high discrepancies especially when corresponding to precursors at trace level. Therefore the assessment of a radionuclide inventory only based on few number of radiochemical measurements leads in most of cases to a gross over or under-estimation that can be detrimental to graphite waste management. A reverse method using an identification calculation-measurement process is proposed in order to assess the radionuclide inventory as precisely as possible.

  16. Carbowaste: treatment and disposal of irradiated graphite and other carbonaceous waste

    International Nuclear Information System (INIS)

    Von Lensa, W.; Rizzato, C.; Baginski, K.; Banford, A.W.; Bradbury, D.; Goodwin, J.; Grambow, B.; Grave, M.J.; Jones, A.N.; Laurent, G.; Pina, G.; Vulpius, D.

    2014-01-01

    The European Project on 'Treatment and Disposal of Irradiated Graphite and other Carbonaceous Waste (CARBOWASTE)' addressed the retrieval, characterization, treatment, reuse and disposal of irradiated graphite with the following main results: - I-graphite waste features significantly depend on the specific manufacture process, on the operational conditions in the nuclear reactor (neutron dose, atmosphere, temperature etc.) and on radiolytic oxidation leading to partial releases of activation products and precursors during operation. - The neutron activation process generates significant recoil energies breaking pre-existing chemical bonds resulting in dislocations of activation products and new chemical compounds. - Most activation products exist in different chemical forms and at different locations. - I-graphite can be partly purified by thermal and chemical treatment processes leaving more leach-resistant waste products. - Leach tests and preliminary performance analyses show that i-graphite can be safely disposed of in a wide range of disposal systems, after appropriate treatment and/or conditioning. (authors)

  17. Heat-to-heat variability of irradiation creep and swelling of HT9 irradiated to high neutron fluence at 400-600{degrees}C

    Energy Technology Data Exchange (ETDEWEB)

    Toloczko, M.B.; Garner, F.A. [Pacific Northwest National Lab., Richland, WA (United States)

    1996-10-01

    Irradiation creep data on ferritic/martensitic steels are difficult and expensive to obtain, and are not available for fusion-relevant neutron spectra and displacement rates. Therefore, an extensive creep data rescue and analysis effort is in progress to characterize irradiation creep of ferritic/martensitic alloys in other reactors and to develop a methodology for applying it to fusion applications. In the current study, four tube sets constructed from three nominally similar heats of HT9 subjected to one of two heat treatments were constructed as helium-pressurized creep tubes and irradiated in FFTF-MOTA at four temperatures between 400 and 600{degrees}C. Each of the four heats exhibited a different stress-free swelling behavior at 400{degrees}C, with the creep rate following the swelling according to the familiar B{sub o} + DS creep law. No stress-free swelling was observed at the other three irradiation temperatures. Using a stress exponent of n = 1.0 as the defining criterion, {open_quotes}classic{close_quotes} irradiation creep was found at all temperatures, but, only over limited stress ranges that decreased with increasing temperature. The creep coefficient B{sub o} is a little lower ({approx}50%) than that observed for austenitic steel, but the swelling-creep coupling coefficient D is comparable to that of austenitic steels. Primary transient creep behavior was also observed at all temperatures except 400{degrees}C, and thermal creep behavior was found to dominate the deformation at high stress levels at 550 and 600{degrees}C.

  18. Irradiated graphite studies prior to decommissioning of G1, G2 and G3 reactors

    International Nuclear Information System (INIS)

    Bonal, J.P.; Vistoli, J.Ph.; Combes, C.

    2005-01-01

    G1 (46 MW th ), G2 (250 MW th ) and G3 (250 MW th ) are the first French plutonium production reactors owned by CEA (Commissariat a l'Energie Atomique). They started to be operated in 1956 (G1), 1959 (G2) and 1960 (G3); their final shutdown occurred in 1968, 1980 and 1984 respectively. Each reactor used about 1200 tons of graphite as moderator, moreover in G2 and G3, a 95 tons graphite wall is used to shield the rear side concrete from neutron irradiation. G1 is an air cooled reactor operated at a graphite temperature ranging from 30 C to 230 C; G2 and G3 are CO 2 cooled reactors and during operation the graphite temperature is higher (140 C to 400 C). These reactors are now partly decommissioned, but the graphite stacks are still inside the reactors. The graphite core radioactivity has decreased enough so that a full decommissioning stage may be considered. Conceming this decommissioning, the studies reported here are: (i) stored energy in graphite, (ii) graphite radioactivity measurements, (iii) leaching of radionuclide ( 14 C, 36 Cl, 63 Ni, 60 Co, 3 H) from graphite, (iv) chlorine diffusion through graphite. (authors)

  19. Irradiation Creep of Ferritic-Martensitic Steels EP-450, EP-823 and EI-852 Irradiated in the BN-350 Reactor over Wide Ranges of Irradiation Temperature and Dose

    International Nuclear Information System (INIS)

    Porollo, S.I.; Konobeev, Y.V.; Ivanov, A.A.; Shulepin, S.V.; Garner, F.

    2007-01-01

    Full text of publication follows: Ferritic/martensitic (F/M) steels appear to be the most promising materials for advanced nuclear systems, especially for fusion reactors. Their main advantages are higher resistance to swelling and lower irradiation creep rate as has been repeatedly demonstrated in examinations of these materials after irradiation. Nevertheless, available experimental data on irradiation resistance of F/M steels are insufficient, with the greatest deficiency of data for high doses and for both low and high irradiation temperatures. From the very beginning of operation the BN-350 fast reactor has been used for irradiation of specimens of structural materials, including F/M steels. The most unique feature of BN-350 was its low inlet sodium temperature, allowing irradiation at temperatures over a very wide range of temperatures compared with the range in other fast reactors. In this paper data are presented on swelling and irradiation creep of three Russian F/M steels EP-450, EP-823 and EI-852, irradiated in experimental assemblies of the BN-350 reactor at temperatures in the range of 305-700 deg. C to doses ranging from 20 to 89 dpa. The investigation was performed using gas-pressurized creep tubes with hoop stresses in the range of 0 - 294 MPa. (authors)

  20. Accelerator-Based Irradiation Creep of Pyrolytic Carbon Used in TRISO Fuel Particles for the (VHTR) Very High Temperature Reactors

    International Nuclear Information System (INIS)

    Wang, Lumin; Was, Gary

    2010-01-01

    Pyrolytic carbon (PyC) is one of the important structural materials in the TRISO fuel particles which will be used in the next generation of gas-cooled very-high-temperature reactors (VHTR). When the TRISO particles are under irradiation at high temperatures, creep of the PyC layers may cause radial cracking leading to catastrophic particle failure. Therefore, a fundamental understanding of the creep behavior of PyC during irradiation is required to predict the overall fuel performance.

  1. Processing of Irradiated Graphite to Meet Acceptance Criteria for Waste Disposal. Results of a Coordinated Research Project

    International Nuclear Information System (INIS)

    2016-05-01

    Graphite is widely used in the nuclear industry and in research facilities and this has led to increasing amounts of irradiated graphite residing in temporary storage facilities pending disposal. This publication arises from a coordinated research project (CRP) on the processing of irradiated graphite to meet acceptance criteria for waste disposal. It presents the findings of the CRP, the general conclusions and recommendations. The topics covered include, graphite management issues, characterization of irradiated graphite, processing and treatment, immobilization and disposal. Included on the attached CD-ROM are formal reports from the participants

  2. Investigation of reactivity between SiC and Nb-1Zr in planned irradiation creep experiments

    Energy Technology Data Exchange (ETDEWEB)

    Lewinsohn, C.A.; Hamilton, M.L.; Jones, R.H.

    1997-08-01

    Thermodynamic calculations and diffusion couple experiments showed that SiC and Nb-1Zr were reactive at the upper range of temperatures anticipated in the planned irradiation creep experiment. Sputter-deposited aluminum oxide (Al{sub 2}O{sub 3}) was selected as a diffusion barrier coating. Experiments showed that although the coating coarsened at high temperature it was an effective barrier for diffusion of silicon from SiC into Nb-1Zr. Therefore, to avoid detrimental reactions between the SiC composite and the Nb-1Zr pressurized bladder during the planned irradiation creep experiment, a coating of Al{sub 2}O{sub 3} will be required on the Nb-1Zr bladder.

  3. Development of an apparatus for measuring the thermal conductivity of irradiated or non-irradiated graphite

    International Nuclear Information System (INIS)

    Bocquet, M.; Micaud, G.

    1962-01-01

    An apparatus was developed for measuring the thermal conductivity coefficient K of irradiated or non-irradiated graphite. The measurement of K at around room temperature with an accuracy of about 6% is possible. The study specimen is placed in a vacuum between a hot and a cold source which create a temperature gradient ΔΘ/ Δx in the steady state. The amount of heat transferred, Q, is deduced from the electrical power dissipated at the hot source, after allowing for heat losses. The thermal conductivity coefficient is defined as: K = Q/S. Δx/ΔΘ, S being the cross section of the sample. Systematic studies have made it possible to determine the mean values of the thermal conductivity. (authors) [fr

  4. Development of out-of-pile version of instrumented irradiation capsule for determination of online creep deformation

    International Nuclear Information System (INIS)

    Venkatesu, Sadu; Saxena, Rajesh; Chaurasia, P.K.; Muthuganesh, M.; Murugan, S.; Venugopal, S.

    2016-01-01

    Materials used for fuel cladding and structural components in fast reactors can undergo significant dimensional and physical changes due to exposure to high energy neutrons. At high temperatures in nuclear environment, material undergoes considerable deformation due to thermal and irradiation creep. Diametral increase of fuel pin due to thermal and irradiation creep, apart from irradiation swelling, reduces the coolant flow area around the fuel pins affecting the effective removal of heat generated in the fuel pins. The changes due to creep can be determined by two types of material irradiation tests in reactor. The first type includes non-instrumented irradiation tests with specimen dimensional evaluations carried out in post-irradiation examinations. The second type includes instrumented irradiation tests with online monitoring and/or controlling of test conditions and real time measurement of changes in dimensions of the specimen. During instrumented irradiation tests, parameters such as specimen temperature, the load exerted on the specimen, specimen elongation, etc. can be monitored and/or controlled using suitable components such as linear variable differential transformers (LVDTs), bellows, thermocouples, etc. Instrumented irradiation experiments in reactors are relatively complex in design but can provide full information on the experimental parameters. Such benefits provide motivation for development of instrumented irradiation capsule to measure creep behavior online during in-pile instrumented irradiation tests. Out-of-pile version of the instrumented irradiation capsule for determination of online creep deformation has been developed and tested in the furnace by raising the temperature gradually up to 330 °C. This paper discusses the details of the design, assembly of experimental set up and experimental results of the out-of-pile version of instrumented capsule developed in our laboratory for determination of online creep deformation. (author)

  5. Irradiation test plan of oxidation-resistant graphite in WWR-K Research Reactor

    International Nuclear Information System (INIS)

    Sumita, Junya; Shibata, Taiju; Sakaba, Nariaki; Osaki, Hirotaka; Kato, Hideki; Fujitsuka, Kunihiro; Muto, Takenori; Gizatulin, Shamil; Shaimerdenov, Asset; Dyussambayev, Daulet; Chakrov, Petr

    2014-01-01

    Graphite materials are used for the in-core components of High Temperature Gas-cooled Reactor (HTGR) which is a graphite-moderated and helium gas-cooled reactor. In the case of air ingress accident in HTGR, SiO_2 protective layer is formed on the surface of SiC layer in TRISO CFP and oxidation of SiC does not proceed and fission products are retained inside the fuel particle. A new safety concept for the HTGR, called Naturally Safe HTGR, has been recently proposed. To enhance the safety of Naturally Safe HTGR ultimately, it is expected that oxidation-resistant graphite is used for graphite components to prevent the TRISO CFPs and fuel compacts from failure. SiC coating is one of candidate methods for oxidation-resistant graphite. JAEA and four graphite companies launched R&Ds to develop the oxidation-resistant graphite and the International Science and Technology Center (ISTC) partner project with JAEA and INP was launched to investigate the irradiation effects on the oxidation-resistant graphite. To determine grades of the oxidation-resistant graphite which will be adopted as irradiation test, a preliminary oxidation test was carried out. This paper described the results of the preliminary oxidation test, the plan of out-of-pile test, irradiation test and post-irradiation test (PIE) of the oxidation-resistant graphite. The results of the preliminary oxidation test showed that the integrity of the oxidation resistant graphite was confirmed and that all of grades used in the preliminary test can be adopted as the irradiation test. Target irradiation temperature was determined to be 1473 (K) and neutron fluence was determined to be from 0.54 × 10"2"5through 1.4 × 10"2"5 (/m"2, E>0.18MeV). Weight change, oxidation rate, activation energy, surface condition, etc. will be evaluated in out-of-pile test and weight change, irradiation effect on oxidation rate and activation energy, surface condition, etc. will be evaluated in PIE. (author)

  6. AGC-2 Graphite Preirradiation Data Analysis Report

    Energy Technology Data Exchange (ETDEWEB)

    William Windes; W. David Swank; David Rohrbaugh; Joseph Lord

    2013-08-01

    This report described the specimen loading order and documents all pre-irradiation examination material property measurement data for the graphite specimens contained within the second Advanced Graphite Capsule (AGC-2) irradiation capsule. The AGC-2 capsule is the second in six planned irradiation capsules comprising the Advanced Graphite Creep (AGC) test series. The AGC test series is used to irradiate graphite specimens allowing quantitative data necessary for predicting the irradiation behavior and operating performance of new nuclear graphite grades to be generated which will ascertain the in-service behavior of the graphite for pebble bed and prismatic Very High Temperature Reactor (VHTR) designs. Similar to the AGC-1 specimen pre-irradiation examination report, material property tests were conducted on specimens from 18 nuclear graphite types but on an increased number of specimens (512) prior to loading into the AGC-2 irradiation assembly. All AGC-2 specimen testing was conducted at Idaho National Laboratory (INL) from October 2009 to August 2010. This report also details the specimen loading methodology for the graphite specimens inside the AGC-2 irradiation capsule. The AGC-2 capsule design requires “matched pair” creep specimens that have similar dose levels above and below the neutron flux profile mid-plane to provide similar specimens with and without an applied load. This document utilized the neutron flux profile calculated for the AGC-2 capsule design, the capsule dimensions, and the size (length) of the selected graphite and silicon carbide samples to create a stacking order that can produce “matched pairs” of graphite samples above and below the AGC-2 capsule elevation mid-point to provide specimens with similar neutron dose levels.

  7. Tests for removal of Co-60 and Eu-154 from irradiated graphite in the TRIGA Reactor

    International Nuclear Information System (INIS)

    Arsene, Carmen

    2009-01-01

    The irradiated graphite in Romania is mainly generated in the thermal columns of TRIGA and WWER-S research reactors (about 9 tones). It was found that the radionuclide content of the graphite irradiated in the TRIGA research reactor is mainly due to C-14 (103 Bq/g), Eu-152 (600-700 Bq/g) and Co-60 (130-150 Bq/g) and low amounts of Eu-154 and Cs-137, depending on location in the thermal column and on irradiation history. In order to minimize the waste inventory and volume in view of their final disposal, in the present paper we show the results of experiments performed for developing and optimizing methods for the chemical decontamination of the irradiated graphite. These procedures are based on strong alkaline solutions for Eu-152 and strong acid solutions for Co-60. The influence of the process parameters on the decontamination factor is investigated. (authors)

  8. Temperature dependence of the thermal expansion of neutron-irradiated pyrolytic carbon and graphite

    International Nuclear Information System (INIS)

    Matsuo, Hideto

    1988-01-01

    The effects of neutron irradiation and annealing on the temperature dependence of the linear thermal expansion of pyrolytic carbon and graphite were investigated after irradiation at 930-1280 0 C to a maximum neutron fluence of 2.84 x 10 25 m -2 (E > 29 fJ). After irradiation, little change in the thermal expansion of pyrolytic graphite was observed. However, as-deposited pyrolytic carbon showed an increase in thermal expansion in the perpendicular direction, a decrease in the direction parallel to the deposition plane, and also an increase in the anisotropy of the thermal expansion. Annealing at 2000 0 C did not cause any effective changes for irradiated specimens of either as-deposited pyrolytic carbon or pyrolytic graphite. (author)

  9. Effects of impurity trapping on irradiation-induced swelling and creep

    Energy Technology Data Exchange (ETDEWEB)

    Mansur, L. K.; Yoo, M. H.

    1977-12-01

    A general theory of the effects of point defect trapping on radiation-induced swelling and creep deformation rates is developed. The effects on the fraction of defects recombining, and on void nucleation, void growth and creep due to the separate processes of dislocation climb-glide and dislocation climb (the so-called SIPA mechanism) are studied. Trapping of vacancies or interstitials increases total recombination and decreases the rates of deformation processes. For fixed trapping parameters, the reduction is largest for void nucleation, less for void growth and creep due to dislocation climb-glide, and least for creep due to dislocation climb. With this formation, the effects of trapping at multiple vacancy and interstitial traps and of spatial and temporal variation in trap concentrations may be determined. Alternative pictures for viewing point defect trapping in terms of effective recombination and diffusion coefficients are derived. It is shown that previous derivations of these coefficients are incorrect. A rigorous explanation is given of the well-known numerical result that interstitial trapping is significant only if the binding energy exceeds the difference between the vacancy and interstitial migration energies, while vacancy trapping is significant even at small binding energies. Corrections which become necessary at solute concentrations above about 0.1% are described. Numerical results for a wide range of material and irradiation parameters are presented.

  10. A discussion of possible mechanisms affecting fission product transport in irradiated and unirradiated nuclear grade graphite

    International Nuclear Information System (INIS)

    Firth, M.J.

    1977-09-01

    137 Cs, 85 Sr, and sup(110m)Ag adsorption experiments were conducted on three graphite powders with differing amounts of specific basal and edge surface areas. No direct proportionality was found between the specific amounts of the isotopes adsorbed and either of the surface characteristics. There appears to be some correlation with the specific basal surface area despite the fact that each isotope behaves differently. Factors that might influence the adsorption behaviour of Cs and Ag during reactor irradiation and heat treatment of nuclear grade graphites are discussed. These include the form of Cs with the graphite surface. A model based on Cs adsorption at vacancy clusters is used to analyse adsorption experiments. A possible explanation for the behaviour of Ag through the migration of graphite impurities from the bulk of the graphite to the pore surface is also discussed. (author)

  11. Change in physical properties of high density isotropic graphites irradiated in the ?JOYO? fast reactor

    Science.gov (United States)

    Maruyama, T.; Kaito, T.; Onose, S.; Shibahara, I.

    1995-08-01

    Thirteen kinds of isotropic graphites with different density and maximum grain size were irradiated in the experimental fast reactor "JOYO" to fluences from 2.11 to 2.86 × 10 26 n/m 2 ( E > 0.1 MeV) at temperatures from 549 to 597°C. Postirradiation examination was carried out on the dimensional changes, elastic modulus, and thermal conductivity of these materials. Dimensional change results indicate that the graphites irradiated at lower fluences showed shrinkage upon neutron irradiation followed by increase with increasing neutron fluences, irrespective of differences in material parameters. The Young's modulus and Poisson's ratio increased by two to three times the unirradiated values. The large scatter found in Poisson's ratio of unirradiated materials became very small and a linear dependence on density was obtained after irradiation. The thermal conductivity decreased to one-fifth to one-tenth of unirradiated values, with a negligible change in specific heat. The results of postirradiation examination indicated that the changes in physical properties of high density, isotropic graphites were mainly dominated by the irradiation condition rather than their material parameters. Namely, the effects of irradiation induced defects on physical properties of heavily neutron-irradiated graphites are much larger than that of defects associated with as-fabricated specimens.

  12. Change in physical properties of high density isotropic graphites irradiated in the ''JOYO'' fast reactor

    International Nuclear Information System (INIS)

    Maruyama, T.; Kaito, T.; Onose, S.; Shibahara, I.

    1995-01-01

    Thirteen kinds of isotropic graphites with different density and maximum grain size were irradiated in the experimental fast reactor ''JOYO'' to fluences from 2.11 to 2.86x10 26 n/m 2 (E>0.1 MeV) at temperatures from 549 to 597 C. Postirradiation examination was carried out on the dimensional changes, elastic modulus, and thermal conductivity of these materials. Dimensional change results indicate that the graphites irradiated at lower fluences showed shrinkage upon neutron irradiation followed by increase with increasing neutron fluences, irrespective of differences in material parameters. The Young's modulus and Poisson's ratio increased by two to three times the unirradiated values. The large scatter found in Poisson's ratio of unirradiated materials became very small and a linear dependence on density was obtained after irradiation. The thermal conductivity decreased to one-fifth to one-tenth of unirradiated values, with a negligible change in specific heat. The results of postirradiation examination indicated that the changes in physical properties of high density, isotropic graphites were mainly dominated by the irradiation condition rather than their material parameters. Namely, the effects of irradiation induced defects on physical properties of heavily neutron-irradiated graphites are much larger than that of defects associated with as-fabricated specimens. (orig.)

  13. A systematic study of acoustic emission from nuclear graphites

    International Nuclear Information System (INIS)

    Neighbour, G.B.; McEnaney, B.

    1996-01-01

    Acoustic emission (AE) monitoring has been identified as a possible method to determine internal stresses in nuclear graphites using the Kaiser effect, i.e., on stressing a graphite that has been subject to a prior stress, the onset of AE occurs at the previous peak stress. For three nuclear graphites (PGA, IM1-24 and VNEC), AE was monitored during both monotonic and cyclic loading to failure in tensile, compressive and flexural test modes. For unirradiated graphites, the Kaiser effect was not found in cyclic loading, but a Felicity effect was observed, i.e., the onset of AE occurred below the previously applied peak stress. The Felicity effect was attributed to time-dependent relaxation and recovery processes and was characterized using a new parameter, the Recovery ratio. It was shown that AE can be used to monitor creep strain and creep recovery in graphites at zero load. The AE-time responses from these experiments were fitted to equations similar to those used for creep strain-time at elevated temperatures. The number of AE counts from irradiated graphites were greater than those from unirradiated graphites, subject to similar stresses, due to increases in porosity caused by radiolytic oxidation. A Felicity effect was also observed on cyclic loading of irradiated graphites, but no evidence for a Kaiser effect was found for irradiated graphites loaded monotonically to failure. Thus internal stresses in irradiated graphites could not be measured using AE. This was attributed to relaxation and recovery processes that occur between removing the irradiated graphite from the reactor and AE testing. This work indicated that AE monitoring is not a suitable technique for measuring internal stresses in irradiated graphite. (author). 19 refs, 6 figs, 6 tabs

  14. Exercise of Intercomparison on characterization of graphite irradiated: CW-RRT

    International Nuclear Information System (INIS)

    Pina, G.; Rodriguez, M.; Lara, E.; Magro, E.

    2014-01-01

    Knowledge of inventory of irradiated graphite (i-graphite) of nuclear reactors is a critical parameter for the proper management of waste. For this reason it is of paramount importance have access to reliable analytical methodologies to provide accurate results through radiochemical procedures specifics. La reliability of the result depends on both its precision and the accuracy obtained. this requires specific tasks compared to other methods, such as the extent of an array reference or conducting intercomparison exercises between laboratories. (Author)

  15. Special graphites; Graphites speciaux

    Energy Technology Data Exchange (ETDEWEB)

    Leveque, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    A large fraction of the work undertaken jointly by the Commissariat a l'Energie Atomique (CEA) and the Pechiney Company has been the improvement of the properties of nuclear pile graphite and the opening up of new fields of graphite application. New processes for the manufacture of carbons and special graphites have been developed: forged graphite, pyro-carbons, high density graphite agglomeration of graphite powders by cracking of natural gas, impervious graphites. The physical properties of these products and their reaction with various oxidising gases are described. The first irradiation results are also given. (authors) [French] Ameliorer les proprietes du graphite nucleaire pour empilements et ouvrir de nouveaux domaines d'application au graphite constituent une part importante de l'effort entrepris en commun par le Commissariat a l'Energie Atomique (CEA) et la compagnie PECHINEY. Des procedes nouveaux de fabrication de carbones et graphites speciaux ont ete mis au point: graphite forge, pyrocarbone, graphite de haute densite, agglomeration de poudres de graphite par craquage de gaz naturel, graphites impermeables. Les proprietes physiques de ces produits ainsi que leur reaction avec differents gaz oxydants sont decrites. Les premiers resultats d'irradiation sont aussi donnes. (auteurs)

  16. Source Term Analysis of the Irradiated Graphite in the Core of HTR-10

    Directory of Open Access Journals (Sweden)

    Xuegang Liu

    2017-01-01

    Full Text Available The high temperature gas-cooled reactor (HTGR has potential utilization due to its featured characteristics such as inherent safety and wide diversity of utilization. One distinct difference between HTGR and traditional pressurized water reactor (PWR is the large inventory of graphite in the core acting as reflector, moderator, or structure materials. Some radionuclides will be generated in graphite during the period of irradiation, which play significant roles in reactor safety, environmental release, waste disposal, and so forth. Based on the actual operation of the 10 MW pebble bed high temperature gas-cooled reactor (HTR-10 in Tsinghua University, China, an experimental study on source term analysis of the irradiated graphite has been done. An irradiated graphite sphere was randomly collected from the core of HTR-10 as sample in this study. This paper focuses on the analytical procedure and the establishment of the analytical methodology, including the sample collection, graphite sample preparation, and analytical parameters. The results reveal that the Co-60, Cs-137, Eu-152, and Eu-154 are the major γ contributors, while H-3 and C-14 are the dominating β emitting nuclides in postirradiation graphite material of HTR-10. The distribution profiles of the above four nuclides are also presented.

  17. Irradiation behavior of bonded structures: impact of stress-enhanced swelling on irradiation creep and elastic properties

    International Nuclear Information System (INIS)

    Hassan, M.H.; Blanchard, J.P.; Kulcinski, G.L.

    1992-01-01

    The objective of this work is to understand the factors that govern the adhesion of coatings on fusion reactor first walls which are subjected to neutron irradiation. Radiation damage will be a major key point in the design of the many duplex components in fusion reactors. There is a substantial amount of available data showing that stress plays a major role in the onset, and possibly the rate, of void growth in austenitic stainless steels. There is also strong support models which predict a coupling of swelling and creep through the stress environment. A parametric study for evidence to stress-enhanced swelling and its connection to creep is conducted for a typical fusion power demonstration reactor. Since microstructural changes are known to affect elastic moduli, the impact of stress enhanced swelling on these moduli are also evaluated

  18. Assessment of different mechanisms of C-14 production in irradiated graphite of RBMK-1500 reactors

    International Nuclear Information System (INIS)

    Narkunas, Ernestas; Smaizys, Arturas; Poskas, Povilas; Kilda, Raimondas

    2010-01-01

    Two RBMK-1500 water-cooled graphite-moderated channel-type power reactors at the Ignalina Nuclear Power Plant (INPP) are under decommissioning now. The total mass of irradiated graphite in the cores of both units is more than 3600 tons. The main source of uncertainty in the numerical assessment of graphite activity is the uncertainty of the initial impurities content in graphite. Nitrogen is one of the most important impurities, having a large neutron capture cross-section. This impurity may become the dominant source of C-14 production. RBMK reactors graphite stacks operate in the cooling mixture of helium-nitrogen gases and this may additionally increase the quantity of the nitrogen impurity. In this paper the results of the numerical modelling of graphite activation for the INPP Unit I reactor are presented. In order to evaluate the C-14 activity dependence on the nitrogen impurity content, several cases with different nitrogen content were modelled taking into account initial nitrogen impurity quantities in the graphite matrix and possible nitrogen quantities entrapped in the graphite pores from cooling gases. (orig.)

  19. Development of an apparatus for measuring the thermal conductivity of irradiated or non-irradiated graphite; Realisation d'un appareil de mesure de la conductibilite thermique du graphite irradie ou non irradie

    Energy Technology Data Exchange (ETDEWEB)

    Bocquet, M; Micaud, G

    1962-07-01

    An apparatus was developed for measuring the thermal conductivity coefficient K of irradiated or non-irradiated graphite. The measurement of K at around room temperature with an accuracy of about 6% is possible. The study specimen is placed in a vacuum between a hot and a cold source which create a temperature gradient {delta}{theta}/ {delta}x in the steady state. The amount of heat transferred, Q, is deduced from the electrical power dissipated at the hot source, after allowing for heat losses. The thermal conductivity coefficient is defined as: K = Q/S. {delta}x/{delta}{theta}, S being the cross section of the sample. Systematic studies have made it possible to determine the mean values of the thermal conductivity. (authors) [French] Un appareil de mesure du coefficient de conductibilite thermique K du graphite irradie ou non irradie a ete realise. Utilisant le principe du transfert de chaleur, il permet de mesurer K au voisinage de la temperature ambiante avec une precision de 6 pour cent environ. L'echantillon de graphite etudie est place sous vide entre une source chaude et une source froide qui creent en regime permanent un gradient de temperature {delta}{theta}/{delta}x La quantite de chaleur transferee Q est deduite de la puissance electrique dissipee dans la source chaude en deduisant les pertes thermiques. Le coefficient de conductibilite thermique est defini par: K = Q/S. {delta}x/{delta}{theta} S designant la section de l'echantillon. Des etudes systematiques ont permis de determiner pour differents graphites non irradies les valeurs moyennes des coefficients de conductibilite thermique. Ces etudes ont mis en evidence pour un type de graphite donne, l'influence de la densite apparente sur le coefficient de conductibilite thermique. A partir de mesures effectuees sur des echantillons de graphite irradies preleves par carottage dans les empilements des reacteurs a moderateur de graphite les variations du rapport K0/Ki en fonction de la dose et de la

  20. High-temperature irradiation effects on mechnical properties of HTGR graphites

    International Nuclear Information System (INIS)

    Oku, Tatsuo; Eto, Motokuni; Fujisaki, Katsuo

    1978-04-01

    The irradiation effects on stress-strain relation, Young's modulus, tensile strength, bending strength and compressive strength of HTGR graphites were studied in irradiation temperature ranges of 200 - 300 0 C and 800 - 1400 0 C and in neutron fluences up to 7.4 x 10 20 n/cm 2 and 3 x 10 21 n/cm 2 (> 0.18 MeV). Fracture criteria and strain energy to fracture of the unirradiated and the irradiated graphites were also examined. (1) Neutron fluence dependences are similar in Young's modulus, tensile strength and bending strength. (2) The change of compressive strength and of tensile and bending strengths with neutron fluence differ; the former varies with graphite kind. (3) At lower irradiation temperatures the bending fracture strain energy decreases with increasing neutron fluence and at higher irradiation temperatures it increases. (4) The fracture criteria of graphites deviates from the constant strain energy theory (α = 0.5) and the constant strain theory (α = 1), shifting from α asymptotically equals 0.5 to α asymptotically equals 1 with increasing irradiation temperature. (auth.)

  1. Irradiation creep and void swelling of two LMR heats of HT9 at ∝400 C and 165 dpa

    International Nuclear Information System (INIS)

    Toloczko, M.B.; Garner, F.A.

    1996-01-01

    Two nominally identical heats of HT9 ferritic-martensitic steel were produced, fabricated into pressurized tubes, and then irradiated in FFTF, using identical procedures. After reaching 165 dpa at ∝400 C, small differences in strains associated with both phase-related changes in lattice parameter and void swelling were observed in comparing the two heats. The creep strains, while different, exhibited the same functional dependence on swelling behavior. The derived creep coefficients, the one associated with creep in the absence of swelling and the one directly responsive to swelling, were essentially identical for the two heats. Even more significantly, the creep coefficients for this bcc ferritic-martensitic steel appear to be very similar and possibly identical to those routinely derived from creep experiments on fcc austenitic steels. (orig.)

  2. Irradiation creep and void swelling of two LMR heat of HT9 at ∼400 degrees C and 165 dpa

    International Nuclear Information System (INIS)

    Toloczko, M.B.; Garner, F.A.

    1996-01-01

    Two nominally identical heats of HT9 ferritic-martensitic steel were produced, fabricated into pressurized tubes, and then irradiated in FFTF, using identical procedures. After reaching 165 dpa at ∼400C, small differences in strains associated with both phase-related change in lattice parameter and void swelling were observed in comparing the two heats. The creep strains, while different, exhibited the same functional relationship to the swelling behavior. The derived creep coefficients, the one associated with creep in the absence of swelling and the one directly responsive to swelling, were essentially identical for the two heats. Even more significantly, the creep coefficients for this bcc ferritic-martensitic steel appear to be very similar and possibly identical to those routinely derived from creep experiments on fcc austenitic steels

  3. Evolution of microstructure after irradiation creep in several austenitic steels irradiated up to 120 dpa at 320 °C

    Energy Technology Data Exchange (ETDEWEB)

    Renault-Laborne, A., E-mail: alexandra.renault@cea.fr [DEN-Service de Recherches Métallurgiques Appliquées, CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France); Garnier, J.; Malaplate, J. [DEN-Service de Recherches Métallurgiques Appliquées, CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France); Gavoille, P. [DEN-Service d' Etudes des Matériaux Irradiés, CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France); Sefta, F. [EDF R& D, MMC, Site des Renardières, F-77818, Morêt-sur-Loing Cedex (France); Tanguy, B. [DEN-Service d' Etudes des Matériaux Irradiés, CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France)

    2016-07-15

    Irradiation creep was investigated in different austenitic steels. Pressurized tubes with stresses of 127–220 MPa were irradiated in BOR-60 at 320 °C to 120 dpa. Creep behavior was dependent on both chemical composition and metallurgical state of steels. Different steels irradiated with and without stress were examined by TEM. Without stress, the irradiation produced high densities of dislocation lines and Frank loops and, depending on the type of steels, precipitates. Stress induced an increase of the precipitate mean size and density and, for some grades, an increase of the mean loop size and a decrease of their density. An anisotropy of Frank loop density or size induced by stress was not observed systematically. Dislocation line microstructure seems not to be different between the stressed and unstressed specimens. No cavities were detectable in these specimens. By comparing with the data from this work, the main irradiation creep models are discussed.

  4. Differences in the irradiation effects of IG-110 and IG-430 nuclear graphites : effects of coke difference

    International Nuclear Information System (INIS)

    Chi, Se Hwan; Kim, Gen Chan; Kim, Eung Seon; Hong, Jin Ki; Chang, Jong Hwa

    2005-01-01

    In the high temperature gas cooled reactors (HTGRs), graphite acts as a moderator and reflector as well as a major structural component that may provide channels for the fuel and coolant gas, channels for control and shut down, and thermal and neutron shielding. During a reactor operation, many of the physical, chemical and mechanical properties of these graphite components are significantly modified as a function of the temperature, environment, and an irradiation. On the other hand, currently, all the nuclear graphites are being manufactured from two types of cokes, i.e., petroleum and coal-tar pitch coke, and it has been understood that the type of coke plays the most critical role determining the properties of a specific graphite grade. To investigate the effects of coke types on the irradiation response of a graphite, two graphites of different cokes were irradiated by 3 MeV C+ ions and the differences in the response of ion-irradiation were investigated

  5. Carbon-14 in neutron-irradiated graphite for graphite-moderated reactors. Joint research

    Energy Technology Data Exchange (ETDEWEB)

    Fujii, Kimio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Matsuo, Hideto [Radioactive Waste Management and Nuclear Facility Decommissioning Technology Center, Tokyo (Japan)

    2002-12-01

    The graphite moderated gas cooled reactor operated by the Japan Atomic Power Company was stopped its commercial operation on March 1998, and the decommissioning process has been started. Graphite material is often used as the moderator and the reflector materials in the core of the gas cooled reactor. During the operation, a long life nuclide of {sup 14}C is generated in the graphite by several transmutation reactions. Separation of {sup 14}C isotope and the development of the separation method have been recognized to be critical issues for the decommissioning of the reactor core. To understand the current methodologies for the carbon isotope separation, literature on the subject was surveyed. Also, those on the physical and chemical behavior of {sup 14}C were surveyed. This is because the larger part of the nuclides in the graphite is produced from {sup 14}N by (n,p) reaction, and the location of them in the material tends to be different from those of the other carbon atoms. This report summarizes the result of survey on the open literature about the behavior of {sup 14}C and the separation methods, including the list of the literature on these subjects. (author)

  6. Irradiation damage in graphite. The works of Professor B.T. Kelly

    International Nuclear Information System (INIS)

    Marsden, B.J.

    1996-01-01

    The irradiation damage produced in graphite by energetic neutrons (>100eV) has been extensively studied because of the use of graphite as a moderator in thermal nuclear reactors. In recent times, graphite has been adopted as the protective tiling of the inner wall of experimental fusion systems and property changes due to fusion neutrons have become important. The late Professor B.T. Kelly reviewed the work carried out on the irradiation behaviour of graphite since the 1940s. This work is particularly timely as the scale of research into the effects of fission neutrons has been greatly reduced and many of the active researchers have retired. In recent years, new programmes of work are being formulated for the use of graphite in both the field of high temperature reactor systems and fusion systems. It is therefore important that the knowledge gained by Professor Kelly and other workers is not lost but passed on to future generations of nuclear scientists and engineers. This paper reviews Professor Kelly's last work, it also draws on the experience gained during many long discussions with Brian during the years he worked closely with the present graphite team at AEA Technology. It is hoped to publish his work in full in the near future. (author). 13 refs, 14 figs, 3 tabs

  7. Swift heavy ions induced irradiation effects in monolayer graphene and highly oriented pyrolytic graphite

    International Nuclear Information System (INIS)

    Zeng, J.; Yao, H.J.; Zhang, S.X.; Zhai, P.F.; Duan, J.L.; Sun, Y.M.; Li, G.P.; Liu, J.

    2014-01-01

    Monolayer graphene and highly oriented pyrolytic graphite (HOPG) were irradiated by swift heavy ions ( 209 Bi and 112 Sn) with the fluence between 10 11 and 10 14 ions/cm 2 . Both pristine and irradiated samples were investigated by Raman spectroscopy. It was found that D and D′ peaks appear after irradiation, which indicated the ion irradiation introduced damage both in the graphene and graphite lattice. Due to the special single atomic layer structure of graphene, the irradiation fluence threshold Φ th of the D band of graphene is significantly lower ( 11 ions/cm 2 ) than that (2.5 × 10 12 ions/cm 2 ) of HOPG. The larger defect density in graphene than in HOPG indicates that the monolayer graphene is much easier to be damaged than bulk graphite by swift heavy ions. Moreover, different defect types in graphene and HOPG were detected by the different values of I D /I D′ . For the irradiation with the same electronic energy loss, the velocity effect was found in HOPG. However, in this experiment, the velocity effect was not observed in graphene samples irradiated by swift heavy ions

  8. Proton irradiated graphite grades for a long baseline neutrino facility experiment

    Directory of Open Access Journals (Sweden)

    N. Simos

    2017-07-01

    Full Text Available In search of a low-Z pion production target for the Long Baseline Neutrino Facility (LBNF of the Deep Underground Neutrino Experiment (DUNE four graphite grades were irradiated with protons in the energy range of 140–180 MeV, to peak fluence of ∼6.1×10^{20}  p/cm^{2} and irradiation temperatures between 120–200 °C. The test array included POCO ZXF-5Q, Toyo-Tanso IG 430, Carbone-Lorraine 2020 and SGL R7650 grades of graphite. Irradiation was performed at the Brookhaven Linear Isotope Producer. Postirradiation analyses were performed with the objective of (a comparing their response under the postulated irradiation conditions to guide a graphite grade selection for use as a pion target and (b understanding changes in physical and mechanical properties as well as microstructure that occurred as a result of the achieved fluence and in particular at this low-temperature regime where pion graphite targets are expected to operate. A further goal of the postirradiation evaluation was to establish a proton-neutron correlation damage on graphite that will allow for the use of a wealth of available neutron-based damage data in proton-based studies and applications. Macroscopic postirradiation analyses as well as energy dispersive x-ray diffraction of 200 KeV x rays at the NSLS synchrotron of Brookhaven National Laboratory were employed. The macroscopic analyses revealed differences in the physical and strength properties of the four grades with behavior however under proton irradiation that qualitatively agrees with that reported for graphite under neutrons for the same low temperature regime and in particular the increase of thermal expansion, strength and Young’s modulus. The proton fluence level of ∼10^{20}  cm^{−2} where strength reaches a maximum before it begins to decrease at higher fluences has been identified and it agrees with neutron-induced changes. X-ray diffraction analyses of the proton irradiated graphite

  9. Rapid analysis method for the determination of 14C specific activity in irradiated graphite.

    Directory of Open Access Journals (Sweden)

    Vidmantas Remeikis

    Full Text Available 14C is one of the limiting radionuclides used in the categorization of radioactive graphite waste; this categorization is crucial in selecting the appropriate graphite treatment/disposal method. We propose a rapid analysis method for 14C specific activity determination in small graphite samples in the 1-100 μg range. The method applies an oxidation procedure to the sample, which extracts 14C from the different carbonaceous matrices in a controlled manner. Because this method enables fast online measurement and 14C specific activity evaluation, it can be especially useful for characterizing 14C in irradiated graphite when dismantling graphite moderator and reflector parts, or when sorting radioactive graphite waste from decommissioned nuclear power plants. The proposed rapid method is based on graphite combustion and the subsequent measurement of both CO2 and 14C, using a commercial elemental analyser and the semiconductor detector, respectively. The method was verified using the liquid scintillation counting (LSC technique. The uncertainty of this rapid method is within the acceptable range for radioactive waste characterization purposes. The 14C specific activity determination procedure proposed in this study takes approximately ten minutes, comparing favorably to the more complicated and time consuming LSC method. This method can be potentially used to radiologically characterize radioactive waste or used in biomedical applications when dealing with the specific activity determination of 14C in the sample.

  10. Rapid analysis method for the determination of 14C specific activity in irradiated graphite.

    Science.gov (United States)

    Remeikis, Vidmantas; Lagzdina, Elena; Garbaras, Andrius; Gudelis, Arūnas; Garankin, Jevgenij; Plukienė, Rita; Juodis, Laurynas; Duškesas, Grigorijus; Lingis, Danielius; Abdulajev, Vladimir; Plukis, Artūras

    2018-01-01

    14C is one of the limiting radionuclides used in the categorization of radioactive graphite waste; this categorization is crucial in selecting the appropriate graphite treatment/disposal method. We propose a rapid analysis method for 14C specific activity determination in small graphite samples in the 1-100 μg range. The method applies an oxidation procedure to the sample, which extracts 14C from the different carbonaceous matrices in a controlled manner. Because this method enables fast online measurement and 14C specific activity evaluation, it can be especially useful for characterizing 14C in irradiated graphite when dismantling graphite moderator and reflector parts, or when sorting radioactive graphite waste from decommissioned nuclear power plants. The proposed rapid method is based on graphite combustion and the subsequent measurement of both CO2 and 14C, using a commercial elemental analyser and the semiconductor detector, respectively. The method was verified using the liquid scintillation counting (LSC) technique. The uncertainty of this rapid method is within the acceptable range for radioactive waste characterization purposes. The 14C specific activity determination procedure proposed in this study takes approximately ten minutes, comparing favorably to the more complicated and time consuming LSC method. This method can be potentially used to radiologically characterize radioactive waste or used in biomedical applications when dealing with the specific activity determination of 14C in the sample.

  11. Relationship between swelling and irradiation creep in cold-worked PCA stainless steel irradiated to ∼178 dpa at ∼400 degrees C

    International Nuclear Information System (INIS)

    Toloczko, M.B.

    1993-09-01

    The eighth and final irradiation segment for pressurized tubes constructed from the fusion Prime Candidate Alloy (PCA) has been completed in FFTF. At 178 dpa and ∼400 degrees C, the irradiation creep of 20% cold-worked PCA has become dominated by the open-quotes creep disappearanceclose quotes phenomenon. The total diametral deformation rate has reached the limiting value of 0.33%/dpa at the three highest stress levels employed in this test. The stress-enhancement of swelling tends to camouflage the onset of creep disappearance, however, requiring the use of several non-traditional techniques to extract the creep coefficients. No failures occurred in these tubes, even though the swelling ranged from ∼20 to ∼40%

  12. Relationship between swelling and irradiation creep in cold-worked PCA stainless steel irradiated to similar 178 dpa at similar 400 C

    International Nuclear Information System (INIS)

    Toloczko, M.B.; Garner, F.A.

    1994-01-01

    The eighth and final irradiation segment for pressurized tubes constructed from the fusion Prime Candidate Alloy (PCA) has been completed in FFTF. At 178 dpa and similar 400 C, the irradiation creep of 20% cold-worked PCA has become dominated by the ''creep disappearance'' phenomenon. The total diametral deformation rate has reached the limiting value of 0.33%/dpa at the three highest stress levels employed in this test. The stress-enhancement of swelling tends to camouflage the onset of creep disappearance, however, requiring the use of several non-traditional techniques to extract the creep coefficients. No failures occurred in these tubes, even though the swelling ranged from similar 20 to 40%. ((orig.))

  13. Relationship between swelling and irradiation creep in cold-worked PCA stainless steel irradiated to similar 178 dpa at similar 400 C

    Energy Technology Data Exchange (ETDEWEB)

    Toloczko, M.B. (Department of Chemical and Nuclear Engineering, University of California, Santa Barbara, CA 93106 (United States)); Garner, F.A. (Pacific Northwest Laboratory, Richland, WA 99352 (United States))

    1994-09-01

    The eighth and final irradiation segment for pressurized tubes constructed from the fusion Prime Candidate Alloy (PCA) has been completed in FFTF. At 178 dpa and similar 400 C, the irradiation creep of 20% cold-worked PCA has become dominated by the creep disappearance'' phenomenon. The total diametral deformation rate has reached the limiting value of 0.33%/dpa at the three highest stress levels employed in this test. The stress-enhancement of swelling tends to camouflage the onset of creep disappearance, however, requiring the use of several non-traditional techniques to extract the creep coefficients. No failures occurred in these tubes, even though the swelling ranged from similar 20 to 40%. ((orig.))

  14. Thermal conductivity degradation of graphites due to neutron irradiation at low temperature

    International Nuclear Information System (INIS)

    Snead, L.L.; Burchell, T.D.

    1995-01-01

    Several graphites and carbon/carbon composites (C/C's) have been irradiated with fission neutrons near 150 C and at fluences up to a displacement level of 0.24 dpa. The unirradiated room temperature thermal conductivity of these materials varied from 114 W/m K for H-451 isotropic graphite, to 670 W/m K for a unidirectional FMI-1D C/C composite. At the irradiation temperature a saturation reduction in thermal conductivity was seen to occur at displacement levels of approximately 0.1 dpa. All materials were seen to degrade to approximately 10 to 14% of their original thermal conductivity after irradiation. The significant recovery of thermal conductivity due to post-irradiation isochronal anneals is also presented. (orig.)

  15. Electron spin resonance in neutron-irradiated graphite. Dependence on temperature and effect of annealing; Resonance paramagnetique du graphite irradie aux neutrons. Variation en fonction de la temperature et experiences de recuit

    Energy Technology Data Exchange (ETDEWEB)

    Kester, T [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires, Laboratoire de resonance magnetique

    1967-09-01

    The temperature dependence of the electron spin resonance signal from neutron irradiated graphite has been studied. The results lead to an interpretation of the nature of the paramagnetic centers created by irradiation. In annealing experiments on graphite samples, which had been irradiated at low temperature, two annealing peaks and one anti-annealing peak were found. Interpretations are proposed for these peaks. (author) [French] Le graphite irradie aux neutrons a ete etudie par resonance paramagnetique electronique en fonction de la temperature. La nature des centres paramagnetiques crees par irradiation est interpretee a l'aide des resultats. Des experiences de recuit sur des echantillons de graphite irradie a 77 deg. K ont permis de mettre en evidence deux pics de recuit et un pic d'anti-recuit, pour lesquels des interpretations sont proposees. (auteur)

  16. Role of Defects in Swelling and Creep of Irradiated SiC

    Energy Technology Data Exchange (ETDEWEB)

    Szlufarska, Izabela [Univ. of Wisconsin, Madison, WI (United States); Voyles, Paul [Univ. of Wisconsin, Madison, WI (United States); Sridharan, Kumar [Univ. of Wisconsin, Madison, WI (United States); Katoh, Yutai [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-01-16

    Silicon carbide is a promising cladding material because of its high strength and relatively good corrosion resistance. However, SiC is brittle and therefore SiC-based components need to be carefully designed to avoid cracking and failure by fracture. In design of SiC-based composites for nuclear reactor applications it is essential to take into account how mechanical properties are affected by radiation and temperature, or in other words, what strains and stresses develop in this material due to environmental conditions. While thermal strains in SiC can be predicted using classical theories, radiation-induced strains are much less understood. In particular, it is critical to correctly account for radiation swelling and radiation creep, which contribute significantly to dimensional instability of SiC under radiation. Swelling typically increases logarithmically with radiation dose and saturates at relatively low doses (damage levels of a few dpa). Consequently, swelling-induced stresses are likely to develop within a few months of operation of a reactor. Radiation-induced volume swelling in SiC can be as high as 2%, which is significantly higher than the cracking strain of 0.1% in SiC. Swelling-induced strains will lead to enormous stresses and fracture, unless these stresses can be relaxed via some other mechanism. An effective way to achieve stress relaxation is via radiation creep. Although it has been hypothesized that both radiation swelling and radiation creep are driven by formation of defect clusters, existing models for swelling and creep in SiC are limited by the lack of understanding of specific defects that form due to radiation in the range of temperatures relevant to fuel cladding in light water reactors (LWRs) (<1000°C). For example, defects that can be detected with traditional transmission electron microscopy (TEM) techniques account only for 10-45% of the swelling measured in irradiated SiC. Here, we have undertaken an integrated experimental and

  17. Role of Defects in Swelling and Creep of Irradiated SiC

    International Nuclear Information System (INIS)

    Szlufarska, Izabela; Voyles, Paul; Sridharan, Kumar; Katoh, Yutai

    2016-01-01

    Silicon carbide is a promising cladding material because of its high strength and relatively good corrosion resistance. However, SiC is brittle and therefore SiC-based components need to be carefully designed to avoid cracking and failure by fracture. In design of SiC-based composites for nuclear reactor applications it is essential to take into account how mechanical properties are affected by radiation and temperature, or in other words, what strains and stresses develop in this material due to environmental conditions. While thermal strains in SiC can be predicted using classical theories, radiation-induced strains are much less understood. In particular, it is critical to correctly account for radiation swelling and radiation creep, which contribute significantly to dimensional instability of SiC under radiation. Swelling typically increases logarithmically with radiation dose and saturates at relatively low doses (damage levels of a few dpa). Consequently, swelling-induced stresses are likely to develop within a few months of operation of a reactor. Radiation-induced volume swelling in SiC can be as high as 2%, which is significantly higher than the cracking strain of 0.1% in SiC. Swelling-induced strains will lead to enormous stresses and fracture, unless these stresses can be relaxed via some other mechanism. An effective way to achieve stress relaxation is via radiation creep. Although it has been hypothesized that both radiation swelling and radiation creep are driven by formation of defect clusters, existing models for swelling and creep in SiC are limited by the lack of understanding of specific defects that form due to radiation in the range of temperatures relevant to fuel cladding in light water reactors (LWRs) (<1000°C). For example, defects that can be detected with traditional transmission electron microscopy (TEM) techniques account only for 10-45% of the swelling measured in irradiated SiC. Here, we have undertaken an integrated experimental and

  18. Nanoscale transformation of sp2 to sp3 of graphite by slow highly charged ion irradiation

    International Nuclear Information System (INIS)

    Meguro, T.; Hida, A.; Koguchi, Y.; Miyamoto, S.; Yamamoto, Y.; Takai, H.; Maeda, K.; Aoyagi, Y.

    2003-01-01

    Nanoscale transformation of electronic states by highly charged ion (HCI) impact on graphite surfaces is described. The high potential energy of slow HCI, which induces multiple emission of electrons from the surface, provides a strong modification of the electronic states of the local area upon graphite surfaces. The HCI impact and the subsequent surface treatment either by electron injection from a scanning tunneling microscopy tip or by He-Cd laser irradiation induce a localized transition from sp 2 to sp 3 hybridization in graphite, resulting in the formation of nanoscale diamond-like structures (nanodiamond) at the impact region. From Raman spectroscopic measurements on sp 2 related peaks, it is found that the HCI irradiation creates vacancy complexes in contrast to ions having a lower charge state, which generate single vacancies. It is of interest that a single impact of HCI creates one nanodiamond structure, suggesting potential applications of HCI in nanoscale material processing

  19. Quality assurance for the IAEA International Database on Irradiated Nuclear Graphite Properties

    International Nuclear Information System (INIS)

    Wickham, A.J.; Humbert, D.

    2006-06-01

    Consideration has been given to the process of Quality Assurance applied to data entered into current versions of the IAEA International Database on Irradiated Nuclear Graphite Properties. Originally conceived simply as a means of collecting and preserving data on irradiation experiments and reactor operation, the data are increasingly being utilised for the preparation of safety arguments and in the design of new graphites for forthcoming generations of graphite-moderated plant. Under these circumstances, regulatory agencies require assurances that the data are of appropriate accuracy and correctly transcribed, that obvious errors in the original documentation are either highlighted or corrected, etc., before they are prepared to accept analyses built upon these data. The processes employed in the data transcription are described in this document, and proposals are made for the categorisation of data and for error reporting by Database users. (author)

  20. Neutron irradiation damage of nuclear graphite studied by high-resolution transmission electron microscopy and Raman spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Krishna, R. [Dalton Cumbrian Facility, Dalton Nuclear Institute, The University of Manchester, Westlakes Science & Technology Park, Moor Row, Whitehaven, Cumbria, CA24 3HA (United Kingdom); Jones, A.N., E-mail: Abbie.Jones@manchester.ac.uk [Nuclear Graphite Research Group, School of MACE, The University of Manchester, Manchester, M13 9PL (United Kingdom); McDermott, L.; Marsden, B.J. [Nuclear Graphite Research Group, School of MACE, The University of Manchester, Manchester, M13 9PL (United Kingdom)

    2015-12-15

    Nuclear graphite components are produced from polycrystalline artificial graphite manufacture from a binder and filler coke with approximately 20% porosity. During the operational lifetime, nuclear graphite moderator components are subjected to fast neutron irradiation which contributes to the change of material and physical properties such as thermal expansion co-efficient, young's modulus and dimensional change. These changes are directly driven by irradiation-induced changes to the crystal structure as reflected through the bulk microstructure. It is therefore of critical importance that these irradiation changes and there implication on component property changes are fully understood. This work examines a range of irradiated graphite samples removed from the British Experimental Pile Zero (BEPO) reactor; a low temperature, low fluence, air-cooled Materials Test Reactor which operated in the UK. Raman spectroscopy and high-resolution transmission electron microscopy (HRTEM) have been employed to characterise the effect of increased irradiation fluence on graphite microstructure and understand low temperature irradiation damage processes. HRTEM confirms the structural damage of the crystal lattice caused by irradiation attributed to a high number of defects generation with the accumulation of dislocation interactions at nano-scale range. Irradiation-induced crystal defects, lattice parameters and crystallite size compared to virgin nuclear graphite are characterised using selected area diffraction (SAD) patterns in TEM and Raman Spectroscopy. The consolidated ‘D’peak in the Raman spectra confirms the formation of in-plane point defects and reflected as disordered regions in the lattice. The reduced intensity and broadened peaks of ‘G’ and ‘D’ in the Raman and HRTEM results confirm the appearance of turbulence and disordering of the basal planes whilst maintaining their coherent layered graphite structure. - Highlights: • Irradiated graphite

  1. Neutron irradiation damage of nuclear graphite studied by high-resolution transmission electron microscopy and Raman spectroscopy

    International Nuclear Information System (INIS)

    Krishna, R.; Jones, A.N.; McDermott, L.; Marsden, B.J.

    2015-01-01

    Nuclear graphite components are produced from polycrystalline artificial graphite manufacture from a binder and filler coke with approximately 20% porosity. During the operational lifetime, nuclear graphite moderator components are subjected to fast neutron irradiation which contributes to the change of material and physical properties such as thermal expansion co-efficient, young's modulus and dimensional change. These changes are directly driven by irradiation-induced changes to the crystal structure as reflected through the bulk microstructure. It is therefore of critical importance that these irradiation changes and there implication on component property changes are fully understood. This work examines a range of irradiated graphite samples removed from the British Experimental Pile Zero (BEPO) reactor; a low temperature, low fluence, air-cooled Materials Test Reactor which operated in the UK. Raman spectroscopy and high-resolution transmission electron microscopy (HRTEM) have been employed to characterise the effect of increased irradiation fluence on graphite microstructure and understand low temperature irradiation damage processes. HRTEM confirms the structural damage of the crystal lattice caused by irradiation attributed to a high number of defects generation with the accumulation of dislocation interactions at nano-scale range. Irradiation-induced crystal defects, lattice parameters and crystallite size compared to virgin nuclear graphite are characterised using selected area diffraction (SAD) patterns in TEM and Raman Spectroscopy. The consolidated ‘D’peak in the Raman spectra confirms the formation of in-plane point defects and reflected as disordered regions in the lattice. The reduced intensity and broadened peaks of ‘G’ and ‘D’ in the Raman and HRTEM results confirm the appearance of turbulence and disordering of the basal planes whilst maintaining their coherent layered graphite structure. - Highlights: • Irradiated graphite exhibits

  2. The effect of compressive stress on the Young's modulus of unirradiated and irradiated nuclear graphites

    International Nuclear Information System (INIS)

    Oku, T.; Usui, T.; Ero, M.; Fukuda, Y.

    1977-01-01

    The Young's moduli of unirradiated and high temperature (800 to 1000 0 C) irradiated graphites for HTGR were measured by the ultrasonic method in the direction of applied compressive stress during and after stressing. The Young's moduli of all the tested graphites decreased with increasing compressive stress both during and after stressing. In order to investigate the reason for the decrease in Young's modulus by applying compressive stress, the mercury pore diameter distributions of a part of the unirradiated and irradiated specimens were measured. The change in pore distribution is believed to be associated with structural changes produced by irradiation and compressive stressing. The residual strain, after removing the compressive stress, showed a good correlation with the decrease in Young's modulus caused by the compressive stress. The decrease in Young's modulus by applying compressive stress was considered to be due to the increase in the mobile dislocation density and the growth or formation of cracks. The results suggest, however, that the mechanism giving the larger contribution depends on the brand of graphite, and in anisotropic graphite it depends on the direction of applied stress and the irradiation conditions. (author)

  3. Irradiation creep at temperatures of 400 degrees C and below for application to near-term fusion devices

    International Nuclear Information System (INIS)

    Grossbeck, M.L.; Gibson, L.T.; Mansur, L.K.

    1996-01-01

    To study irradiation creep at 400 degrees C and below, a series of six austenitic stainless steels and two ferritic alloys was irradiated sequentially in two research reactors where the neutron spectrum was tailored to produce a He production rate typical of a fusion device. Irradiation began in the Oak Ridge Research Reactor; and, after an atomic displacement level of 7.4 dpa, the specimens were moved to the High Flux Isotope Reactor for the remainder of the 19 dpa accumulated. Irradiation temperatures of 60, 200, 330, and 400 degrees C were studied with internally pressurized tubes of type 316 stainless steel, PCA, HT 9, and a series of four laboratory heats of: Fe-13.5Cr-15Ni, Fe-13.5Cr-35Ni, Fe-1 3.5Cr-1 W-0.18Ti, and Fe-16Cr. At 330 degrees C, irradiation creep was shown to be linear in fluence and stress. There was little or no effect of cold-work on creep under these conditions at all temperatures investigated. The HT9 demonstrated a large deviation from linearity at high stress levels, and a minimum in irradiation creep with increasing stress was observed in the Fe-Cr-Ni ternary alloys

  4. Use of state variables in the description of irradiation creep and deformation of metals

    International Nuclear Information System (INIS)

    Hart, E.W.; Li, C.Y.

    1976-01-01

    The understanding of the effects of irradiation on metal creep and deformation are not yet satisfactory, owing in part to the limitations on experimentation in radiation environment. Because of such limitations, theoretical considerations must play a strong role. Virtually all of the theoretical considerations currently employed are based on micro-mechanical models for the deformation behavior. The recent theoretical and experimental development of a plastic equation of state for metal deformation has led to the identification of some of the principal micro-mechanisms in phenomenological terms. The role of the individual mechanisms can be related to the state variables of the description, and those variables are directly accessible measurable quantities. This paper explores how irradiation might affect this description. It is shown that the radiation flux and the radiation fluence are expected to affect different components of the equation of state. The resultant description makes considerable use of the information developed in radiation-free environment. 5 fig

  5. Measurement and computation for sag of calandria tube due to irradiation creep in PHWR

    International Nuclear Information System (INIS)

    Son, S. M.; Lee, W. R.; Lee, S. K.; Lee, J. S.; Kim, T. R.; Na, B. K.; Namgung I.

    2003-01-01

    Calandria tubes and Liquid Injection Shutdown System(LISS) tubes in a Pressurized Heavy Water Reactor(PHWR) are to sag due to irradiation creep and growth during plant operation. When the sag of calandria tube becomes bigger, the calandria tube possibly comes in contact with LISS tube crossing beneath the calandria tube. The contact subsequently may cause the damage on the calandria tube resulting in unpredicted outage of the plant. It is therefore necessary to check the gap between the two tubes in order to periodically confirm no contact by using a proper measure during the plant life. An ultrasonic gap measuring probe assembly which can be inserted into two viewing ports of the calandria was developed in Korea and utilized to measure the sags of both tubes in the PHWR. It was found that the centerlines of calandria tubes and liquid injection shutdown system tubes can be precisely detected by ultrasonic wave. The gaps between two tubes were easily obtained from the relative distance of the measured centerline elevations of the tubes. Based on the irradiation creep equation and the measurement data, a computer program to calculate the sags was also developed. With the computer program, the sag at the end of plant life was predicted

  6. Tuning Vortex Creep in Irradiated YBa2Cu3O7-δ Coated Conductors

    Science.gov (United States)

    Eley, Serena; Kihlstrom, Karen; Holleis, Sigrid; Leroux, Maxime; Rupich, Martin; Miller, Dean; Kayani, Asghar; Welp, Ulrich; Kwok, Wai-Kwong; Civale, Leonardo

    YBa2Cu3O7-δ coated conductors (CCs) show non-monotonic changes in the temperature-dependent creep rate, S (T) , due to mixed pinning landscapes comprised of twin boundaries, planar defects, point defects, and nanoparticle precipitates. Notably, in low fields, there is a conspicuous dip in S as T increases from ~20K to ~65K. The source of this dip is poorly understood. Moreover, pinning landscapes that are favorable for high critical currents, Jc, are not necessarily optimal for low S. We have found that, though oxygen irradiation introduces few-nm-sized defects that result in significant increases in Jc, it is detrimental to creep, increasing S (reducing the dip depth) for T > 20K. Understanding the source of this dip is crucial to engineering pinning landscapes that concurrently promote high Jc and low S. To this end, we study changes in S (T) as we tune the ratio of smaller (point to few-nm-sized) defects to larger nanoparticles in an oxygen-irradiated CC by annealing in O2 at 250°C to 600°C. We observe a steady decrease in S (T > 20K) with increasing annealing temperature. This suggests that pre-existing nanoparticle precipitates are likely responsible for the dip in S (T) , and underlines the fact that the effects of defects are not additive, but rather can be competitive.

  7. Curling and closure of graphitic networks under electron-beam irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Ugarte, D [Ecole Polytechnique Federale, Lausanne (Switzerland)

    1992-10-22

    The discovery of buckminsterfullerene (C[sub 60]) and its production in macroscopic quantities has stimulated a great deal of research. More recently, attention has turned towards other curved graphitic networks, such as the giant fullerenes (C[sub n], n > 100) and carbon nanotubes. A general mechanism has been proposed in which the graphitic sheets bend in an attempt to eliminate the highly energetic dangling bonds present at the edge of the growing structure. Here, I report the response of carbon soot particles and tubular graphitic structures to intense electron-beam irradiation in a high-resolution electron microscope; such conditions resemble a high-temperature regime, permitting a degree of structural fluidity. With increased irradiation, there is a gradual reorganization of the initial material into quasi-spherical particles composed of concentric graphitic shells. This lends weight to the nucleation scheme proposed for fullerenes, and moreover, suggests that planar graphite may not be the most stable allotrope of carbon in systems of limited size. (Author).

  8. Characterization of fresh and irradiated domestic nuclear graphite; Karakterizacija neozracenog i ozracenog domaceg nuklearnog grafita

    Energy Technology Data Exchange (ETDEWEB)

    Marinkovic, S; Suznjevic, C; Bogdanovic, R; Gasic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1965-11-15

    This report shows results of testing the quality of domestic impregnated graphite IGSP-05, and irradiated domestic graphite IGSP-01 as well as the new methos of characterization based on graphite oxidation by liquid agent. Systematic measurement of domestic impregnated graphite enabled conclusions related to its quality and further improvement. Domestic graphite is relatively well graphitized and its properties are approaching standard nuclear graphite, although it still shows some deficiencies. Important deficiencies are significant inhomogeneity and low density. The applied impregnation procedure did not improve significantly the quality of graphite, probably because the material which was impregnated had fine pores. To avoid this porosity it would be necessary to use material with higher granulation. Soot which was present in some blocks probably worsened the quality of graphite and caused dispersion of the obtained results. First tests of irradiated domestic graphite IGSP-01 showed that its behaviour does not differ from standard nuclear graphite in case of low doses. It is necessary to test its properties in case of higher neutron doses before drawing final conclusions. The new method of graphite oxidation by the N{sub 2}SO{sub 4} - Ag{sub 2}Cr{sub 2}O{sub 7} mixture which is highly sensitive on the existence of structural defects is based on detecting the oxidation rate of graphite by measuring the pressure of released CO{sub 2}. Application of the method for testing the domestic and American graphite showed that irradiation caused drastic changes of oxidation rates and similar behaviour of both graphite types. U ovom izvestaju su prikazani rezultati ispitivanja kvaliteta domaceg impregnisanog grafita IGSP-05, rezultati ispitivanja ozracenog domaceg grafita IGSP-01 i opisana je nova uvedena metoda karakterizacije zasnovana na oksidaciji grafita tecnim agensom. Sistematsko merenje osobina domaceg impregnisanog grafita je omogucilo donosenje zakljucaka o

  9. Irradiation damage in graphite due to fast neutrons in fission and fusion systems

    International Nuclear Information System (INIS)

    2000-09-01

    Gas cooled reactors have been in operation for the production of electricity for over forty years, encompassing a total of 56 units operated in seven countries. The predominant experience has been with carbon dioxide cooled reactors (52 units), with the majority operated in the United Kingdom. In addition, four prototype helium cooled power plants were operated in the United States and Germany. The United Kingdom has no plans for further construction of carbon dioxide units, and the last helium cooled unit was shutdown in 1990. However, there has been an increasing interest in modular helium cooled reactors during the 1990s as a possible future nuclear option. Graphite is a primary material for the construction of gas cooled reactor cores, serving as a low absorption neutron moderator and providing a high temperature, high strength structure. Commercial gas cooled reactor cores (both carbon dioxide cooled and helium cooled) utilise large quantities of graphite. The structural behaviour of graphite (strength, dimensional stability, susceptibility to cracking, etc.) is a complex function of the source material, manufacturing process, chemical environment, and temperature and irradiation history. A large body of data on graphite structural performance has accumulated from operation of commercial gas cooled reactors, beginning in the 1950s and continuing to the present. The IAEA is supporting a project to collect graphite data and archive it in a retrievable form as an International Database on Irradiated Nuclear Graphite Properties, with limited general access and more detailed access by participating Member States. Because of the large size of the database, the complexity of the phenomena and the number of variables involved, a general understanding of graphite behaviour is essential to the understanding and use of the data

  10. Irradiation Creep and Swelling of Russian Ferritic-Martensitic Steels Irradiated to Very High Exposures in the BN-350 Fast Reactor at 305-335 degrees C

    International Nuclear Information System (INIS)

    Konobeev, Yury V.; Dvoriashin, Alexander M.; Porollo, S.I.; Shulepin, S.V.; Budylkin, N.I.; Mironova, Elena G.; Garner, Francis A.

    2003-01-01

    Russian ferritic/martensitic (F/M) steels EP-450, EP-852 and EP-823 were irradiated in the BN-350 fast reactor in the form of gas-pressurized creep tubes. The first steel is used in Russia for hexagonal wrappers in fast reactors. The other steels were developed for compatibility with Pb-Bi coolants and serve to enhance our understanding of the general behavior of this class of steels. In an earlier paper we published data on irradiation creep of EP-450 and EP-823 at temperatures between 390 and 520C, with dpa levels ranging from 20 to 60 dpa. In the current paper new data on the irradiation creep and swelling of EP-450 and EP-852 at temperatures between 305 and 335C and doses ranging from 61 to 89 dpa are presented. Where comparisons are possible, it appears that these steels exhibit behavior that is very consistent with that of Western steels. Swelling is relatively low at high neutron exposure and confined to temperatures <420C, but may be camouflaged somewhat by precipitation-related densification. These irradiation creep studies confirm that the creep compliance of F/M steels is about one-half that of austenitic steels.

  11. Irradiation-induced dimensional changes of fuel compacts and graphite sleeves of OGL-1 fuel assemblies

    International Nuclear Information System (INIS)

    Hayashi, Kimio; Minato, Kazuo; Kobayashi, Fumiaki; Tobita, Tsutomu; Kikuchi, Teruo; Kurobane, Shiro; Adachi, Mamoru; Fukuda, Kousaku

    1988-06-01

    Experimental data are summarized on irradiation-induced dimensional changes of fuel compacts and graphite sleeves of the first to ninth OGL-1 fuel assemblies. The range of fast-neutron fluence is up to 4 x 10 24 n/m 2 (E > 0.18 MeV); and that of irradiation temperature is 900 - 1400 deg C for fuel compacts and 800 - 1050 deg C for graphite sleeves. The dimensional change of the fuel compacts was shrinkage under these test conditions, and the shrinkage fraction increased almost linearly with fast-neutron fluence. The shrinkage fraction of the fuel compacts was larger by 20 % in the axial direction than in the radial direction. Influence of the irradiation temperature on the dimensional-change behavior of the fuel compacts was not observed clearly; presumably the influence was hidden by scatter of the data because of low level of the fast-neutron fluence and the resultant small dimensional changes. (author)

  12. Creep test under irradiation with thermal gradient for the cylindrical carbon fiber reinforced carbon composite. Interim report on irradiation examinations: 03M-47AS

    International Nuclear Information System (INIS)

    Baba, Shin-ichi; Sawa, Kazuhiro; Yamaji, Masatoshi; Matsui, Yoshinori; Ishihara, Masahiro

    2007-03-01

    The creep test under irradiation with thermal gradient for the cylindrical carbon fiber reinforced carbon composites (c/c composite) are carried out in the Japan Material Testing Reactor (JMTR). This report described 4-items; first item is design/evaluation of the capsule for the irradiation test, second is before irradiation measurements for the residual strain due to manufactured cylindrical c/c composite, and third is also before irradiation measurements of the distance between 2-holes of predrilled in the specimen and last item is examination of analysis for the irradiation creep with thermal gradient by VIENUS Code. The normal creep test is static mechanical load on the specimen in thermal condition, but this creep test under irradiation capsule is thermal stress due to thermal gradient at inside specimen in the thermal condition. Consequently, it is necessary as large as possible thermal gradient in the narrow space of the capsule inside volume. In which the tungsten rod (W-rod) was inserted to the cylindrical c/c composite specimen, for γ-ray heat generation density occurred highly and so maximize the difference temperatures of surface wall between inside and outside wall of the specimen. The measurement method of the deflection due to the irradiation creep of cylindrical c/c composite was adopted as way of ruptured the specimen among the predrilled distance of 2-holes before/after irradiation. Accordingly as the laser dimensional apparatus used to measure the distance between the 2-holes of specimen exactly, easy and untouchable. And also before irradiation measurement of the residual stress due to the manufactured process was estimated by neutron diffraction used Residual Stress Analyzer (RESA) at JRR-3M in JAEA. The irradiation test was finished as total irradiation time, average temperature and neutron dose showed 4189 hours, 873 K and 8.2x10 24 (E>1.0MeV:m -2 ) respectively. The thermal stress was estimated by the difference temperatures of 4

  13. Erosion of CFC, pyrolytic and boronated graphite under short pulsed laser irradiation

    International Nuclear Information System (INIS)

    Kraaij, G.J.; Bakker, J.; Stad, R.C.L. van der

    1992-07-01

    The effect of short pulsed laser irradiation of '0/3' ms and up to 10 MJ/m 2 on different types of carbon base materials is described. These materials are investigated as candidate protection materials for the Plasma Facing Components of NET/ITER. These materials are: carbon fibre composite graphite, pyrolytic graphite and boronated graphite. The volume of the laser induced craters was measured with an optical topographic scanner, and these data are evaluated with a simple model for the erosion. As a results, the enthalpy of ablation is estimated as 30±3 MJ/kg. A comparison is made with finite element numerical calculations, and the effect of lateral heat transfer is estimated using an analytical model. (author). 8 refs., 23 figs., 4 tabs

  14. Simulating Neutron Radiation Damage of Graphite by In-situ Electron Irradiation

    International Nuclear Information System (INIS)

    Mironov, Brindusa E; Freeman, H M; Brydson, R M D; Westwood, A V K; Scott, A J

    2014-01-01

    Radiation damage in nuclear grade graphite has been investigated using transmission electron microscopy (TEM) and electron energy loss spectroscopy (EELS). Changes in the structure on the atomic scale and chemical bonding, and the relationship between each were of particular interest. TEM was used to study damage in nuclear grade graphite on the atomic scale following 1.92×10 8 electrons nm −2 of electron beam exposure. During these experiments EELS spectra were also collected periodically to record changes in chemical bonding and structural disorder, by analysing the changes of the carbon K-edge. Image analysis software from the 'PyroMaN' research group provides further information, based on (002) fringe analysis. The software was applied to the micrographs of electron irradiated virgin 'Pile Grade A' (PGA) graphite to quantify the extent of damage from electron beam exposure

  15. Graphite for high-temperature reactors

    International Nuclear Information System (INIS)

    Hammer, W.; Leushacke, D.F.; Nickel, H.; Theymann, W.

    1976-01-01

    The different graphites necessary for HTRs are being developed, produced and tested within the Federal German ''Development Programme Nuclear Graphite''. Up to now, batches of the following graphite grades have been manufactured and fully characterized by the SIGRI Company to demonstrate reproducibility: pitch coke graphite AS2-500 for the hexagonal fuel elements and exchangeable reflector blocks; special pitch coke graphite ASI2-500 for reflector blocks of the pebble-bed reactor and as back-up material for the hexagonal fuel elements; graphite for core support columns. The material data obtained fulfill most of the requirements under present specifications. Production of large-size blocks for the permanent side reflector and the core support blocks is under way. The test programme covers all areas important for characterizing and judging HTR-graphites. In-pile testing comprises evaluation of the material for irradiation-induced changes of dimensions, mechanical and thermal properties - including behaviour under temperature cycling and creep behaviour - as well as irradiating fuel element segments and blocks. Testing out-of-pile includes: evaluation of corrosion rates and influence of corrosion on strength; strength measurements; including failure criteria. The test programme has been carried out extensively on the AS2-graphite, and the results obtained show that this graphite is suitable as HTGR fuel element graphite. (author)

  16. Irradiation-induced defects in graphite and glassy carbon studied by positron annihilation

    International Nuclear Information System (INIS)

    Hasegawa, M.; Kajino, M.; Kuwahara, H.; Yamaguchi, S.; Kuramoto, E.; Takenaka, M.

    1992-01-01

    ACAR and positron lifetime measurements have been made on, HOPG, isotropic fine-grained graphites, glassy carbons and C 60 /C 70 . HOPG showed a marked bimodal ACAR distribution along the c-axis. By irradiation of 1.0 X 10 19 fast neutrons/cm 2 remarkable narrowing in the ACAR curves and disappearance of the bimodal distribution were observed. Lifetime in HOPG increased from 225 psec to 289 psec (positron-lifetime in vacancies and their small clusters) by the irradiation. The irradiation on isotropic graphites and glassy carbons, however, gave slight narrowing in ACAR curves and decrease in lifetimes (360 psec → 300psec). This suggests irradiation-induced vacancy trapping in crystallites. In C 60 /C 70 powder two lifetime components were detected: τ 1 =177psec, τ 2 =403psec (I 2 =58%). The former is less than the bulk lifetime of HOPG, while the latter being very close to lifetimes in the isotropic graphites and glassy carbons. This and recent 2D-ACAR study of HOPG surface [15] strongly suggest free and defect surface states around ''soccer ball'' cages

  17. The influence of mechanical deformation on the irradiation creep of AISI 316 stainless steel irradiated in the EBR-II and FFTF fast reactors

    International Nuclear Information System (INIS)

    Garner, F.A.; Gilbert, E.R.

    2007-01-01

    Irradiation creep of stainless steels is thought not to be very responsive to material and environmental variables. To test this perception earlier unpublished experiments conducted in the EBR-II reactor on AISI 316 have been analyzed. While swelling is dependent on the cold-work level at 400-480 o C, the post-transient irradiation creep rate, often called the creep compliance B0, is not dependent on cold-work level. If the tube reaches pressures on reactor start-up that generate above-yield stresses in unirradiated steel, then plastic strains occur prior to significant irradiation, but the post-transient strain rate is identical to that of material that did not exceed the yield stress on start-up. It is shown that both stress-free and stress-affected swelling are isotropic and that the Soderberg relationship is maintained. At temperatures above ∼540 o C thermal creep and stored energy begin to assert themselves, with creep rates accelerating with cold-work and becoming non-linear with stress. These results are in agreement with a similar study on titanium-modified 316 steel in FFTF. (author)

  18. Final report on development and operation of instrumented irradiation capsules for creep experiments on nuclear fuels at FR2

    International Nuclear Information System (INIS)

    Haefner, H.E.; Philipp, K.; Blumhofer, M.

    1980-02-01

    The capsule test rig No. 154 removed from FR2 in April 1979 was the last irradiation rig in a long series of creep experiments. The target of the irradiation tests, started exactly ten years ago, was to investigate the creep behaviour of various ceramic nuclear fuels under different in-pile irradiation conditions. An irradiation test rig had been developed for this purpose which allowed the continuous measurement of changes in length of fuel specimens. A total of 28 capsule test rigs each containing two packages of creep specimens have been irradiated in FR2 during this decade. They included 23 specimen stacks of UO 2 , 16 specimen stacks of UO 2 -PuO 2 , 4 specimen stacks of UN, 10 specimen stacks of (U,Pu) C, and 13 reference specimens of molybdenum. Besides the description of the test facility, the report provides above all a survey of the operation data applicable to the specimens and of the operating experience gathered as well as of the findings obtained in post-irradiation examinations. (orig.) [de

  19. Special graphites

    International Nuclear Information System (INIS)

    Leveque, P.

    1964-01-01

    A large fraction of the work undertaken jointly by the Commissariat a l'Energie Atomique (CEA) and the Pechiney Company has been the improvement of the properties of nuclear pile graphite and the opening up of new fields of graphite application. New processes for the manufacture of carbons and special graphites have been developed: forged graphite, pyro-carbons, high density graphite agglomeration of graphite powders by cracking of natural gas, impervious graphites. The physical properties of these products and their reaction with various oxidising gases are described. The first irradiation results are also given. (authors) [fr

  20. Negative pressure and spallation in graphite targets under nano- and picosecond laser irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Belikov, R S; Khishchenko, K V [Joint Institute for High Temperatures, Russian Academy of Sciences, Moscow (Russian Federation); Krasyuk, I K; Semenov, A Yu; Stuchebryukhov, I A [A M Prokhorov General Physics Institute, Russian Academy of Sciences, Moscow (Russian Federation); Rinecker, T; Schoenlein, A [Goethe University Frankfurt am Main (Germany); Rosmej, O N [GSI Helmholtzzentrum für Schwerionenforschung GmbH, Germany, 64291 Darmstadt, Planckstraße, 1 (Germany); Tomut, M [Technische Universität Darmstadt, Germany, 64289 Darmstadt, Karolinenplatz, 5 (Germany)

    2015-05-31

    We present the results of experiments on the spallation phenomena in graphite targets under shock-wave nano- and picosecond irradiation, which have been performed on Kamerton-T (GPI, Moscow, Russia) and PHELIX (GSI, Darmstadt, Germany) laser facilities. In the range of the strain rates of 10{sup 6} – 10{sup 7} s{sup -1}, the data on the dynamic mechanical strength of the material at rapure (spallation) have been for the first time obtained. With a maximal strain rate of 1.4 × 10{sup 7} s{sup -1}, the spall strength of 2.1 GPa is obtained, which constitutes 64% of the theoretical ultimate tensile strength of graphite. The effect of spallation is observed not only on the rear side of the target, but also on its irradiated (front) surface. With the use of optical and scanning electron microscopes, the morphology of the front and rear surfaces of the targets is studied. By means of Raman scattering of light, the graphite structure both on the target front surface under laser exposure and on its rear side in the spall zone is investigated. A comparison of the dynamic strength of graphite and synthetic diamond is performed. (extreme light fields and their applications)

  1. Characterization of {sup 14}C in neutron irradiated NBG-25 nuclear graphite

    Energy Technology Data Exchange (ETDEWEB)

    LaBrier, Daniel, E-mail: labrdani@isu.edu; Dunzik-Gougar, Mary Lou

    2014-05-01

    Recent studies suggest that the highest concentration of {sup 14}C contamination present in reactor-irradiated graphite exists on the surfaces and within near-surface layers. Surface-sensitive analysis techniques (XPS, ToF-SIMS, SEM/EDS and Raman) were employed to determine the chemical nature of {sup 14}C on irradiated NBG-25 (nuclear grade) graphite surfaces. Several {sup 14}C precursor species are identified on the surfaces of irradiated NBG-25; the quantities of these species decrease at sub-surface depths, which further suggests that {sup 14}C formation is predominantly a surface-concentrated phenomenon. The elevated presence of several surface oxide complexes on irradiated NBG-25 surfaces are attributed directly to neutron irradiation. Larger numbers of oxide bonds were found on irradiated NBG-25 surfaces (when compared to unirradiated samples) in the form of interlattice (e.g. ether) and dangling (e.g. carboxylate and ketone) bonds; the quantities of these bond types also decrease with increasing sub-surface depths.

  2. Evaluation of thermal shock strengths for graphite materials using a laser irradiation method

    International Nuclear Information System (INIS)

    Kim, Jae Hoon; Lee, Young Shin; Kim, Duck Hoi; Park, No Seok; Suh, Jeong; Kim, Jeng O.; Il Moon, Soon

    2004-01-01

    Thermal shock is a physical phenomenon that occurs during the exposure to rapidly high temperature and pressure changes or during quenching of a material. The rocket nozzle throat is exposed to combustion gas of high temperature. Therefore, it is important to select suitable materials having the appropriate thermal shock resistance and to evaluate these materials for rocket nozzle design. The material of this study is ATJ graphite, which is the candidate material for rocket nozzle throat. This study presents an experimental method to evaluate the thermal shock resistance and thermal shock fracture toughness of ATJ graphite using laser irradiation. In particular, thermal shock resistance tests are conducted with changes of specimen thickness, with laser source irradiated at the center of the specimen. Temperature distributions on the specimen surface are detected using type K and C thermocouples. Scanning electron microscope (SEM) is used to observe the thermal cracks on specimen surface

  3. Low energy He+ irradiation effect on graphite surface

    International Nuclear Information System (INIS)

    Asari, E.; Nakamura, K.G.; Kitajima, M.; Kawabe, T.

    1992-01-01

    Study on the lattice disordering and the secondary electron emission under low energy (1-5keV) He + irradiation is reported. Real-time Raman measurements show that difference in the observed Raman spectra for different ion energies is due to the difference of the damage depth. The relation between the observed Raman spectrum and the depth profile of lattice damage is discussed. Energy dependence of the secondary electron emission coefficient are also described. (author)

  4. Polymer surfaces graphitization by low-energy He{sup +} ions irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Geworski, A.; Lazareva, I.; Gieb, K.; Koval, Y.; Müller, P. [Department of Physics, Universität Erlangen-Nürnberg, Erwin-Rommel-Str. 1, 91058 Erlangen (Germany)

    2014-08-14

    The electrical and optical properties of surfaces of polyimide and AZ5214e graphitized by low-energy (1 keV) He{sup +} irradiation at different polymer temperatures were investigated. The conductivity of the graphitized layers can be controlled with the irradiation temperature within a broad range and can reach values up to ∼1000 S/cm. We show that the electrical transport in low-conducting samples is governed by thermally activated hopping, while the samples with a high conductivity show a typical semimetallic behavior. The transition from thermally activated to semimetallic conductance governed by the irradiation temperature could also be observed in optical measurements. The semimetallic samples show an unusually high for graphitic materials carrier concentration, which results in a high extinction coefficient in the visible light range. By analyzing the temperature dependence of the conductance of the semimetallic samples, we conclude that the scattering of charge carriers is dominated by Coulomb interactions and can be described by a weak localization model. The transition from a three to two dimensional transport mechanism at low temperatures consistently explains the change in the temperature dependence of the conductance by cooling, observed in experiments.

  5. Study on structural recovery of graphite irradiated with swift heavy ions at high temperature

    Energy Technology Data Exchange (ETDEWEB)

    Pellemoine, F., E-mail: pellemoi@frib.msu.edu [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI 48824 (United States); Avilov, M. [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI 48824 (United States); Bender, M. [Dept. of Materials Research, GSI Helmholtzzentrum für Schwerionenforschung, Planckstr. 1, Darmstadt 64291 (Germany); Ewing, R.C. [Dept. of Geological Sciences, Stanford University, Stanford, CA 94305-2115 (United States); Fernandes, S. [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI 48824 (United States); Lang, M. [Dept. of Nuclear Engineering, University of Tennessee, Knoxville, TN 37996-2300 (United States); Li, W.X. [Dept. of Geological Sciences, Stanford University, Stanford, CA 94305-2115 (United States); Mittig, W. [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI 48824 (United States); National Superconducting Cyclotron Laboratory, Michigan State University, East Lansing, MI 48824-1321 (United States); Schein, M. [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI 48824 (United States); Severin, D. [Dept. of Materials Research, GSI Helmholtzzentrum für Schwerionenforschung, Planckstr. 1, Darmstadt 64291 (Germany); Tomut, M. [Dept. of Materials Research, GSI Helmholtzzentrum für Schwerionenforschung, Planckstr. 1, Darmstadt 64291 (Germany); Laboratory of Magnetism and Superconductivity, National Institute for Materials Physics NIMP, Bucharest (Romania); Trautmann, C. [Dept. of Materials Research, GSI Helmholtzzentrum für Schwerionenforschung, Planckstr. 1, Darmstadt 64291 (Germany); Dept. of Materials Science, Technische Universität Darmstadt, Darmstadt (Germany); and others

    2015-12-15

    Thin graphite foils bombarded with an intense high-energy (8.6 MeV/u) gold beam reaching fluences up to 1 × 10{sup 15} ions/cm{sup 2} lead to swelling and electrical resistivity changes. As shown earlier, these effects are diminished with increasing irradiation temperature. The work reported here extends the investigation of beam induced changes of these samples by structural analysis using synchrotron X-ray diffraction and transmission electron microscope. A nearly complete recovery from swelling at irradiation temperatures above about 1500 °C is identified.

  6. Post-irradiation creep properties of four plates and two forgings DIN 1.4948 steel from the SNR-300 permanent primary structures

    International Nuclear Information System (INIS)

    Schaaf, B. van der.

    1987-01-01

    The safety authorities, involved in the licensing procedure of the SNR-300, have required the determination of the irradiation effect on the heat-to-heat variation of tensile and creep properties of Werkst. No. DIN 1.4948 austenitic stainless steel. These data are lacking in the present codes and they are necessary for the design and safety considerations of the permanent structures. Results are presented of about 200 tests on irradiated and unirradiated material of 6 heats used in the production of the SNR-300 permanent structures. After irradiation in the HFR-Petten to neutron fluences relevant for the SNR-300 service conditions post-irradiation tensile and creep tests (up to 10,000 hrs rupture time) were performed in the temperature range 723 K to 923 K. All heats are embrittled by irradiation resulting in reduction of rupture times, creep strength and ultimate tensile strength. The considerable reduction is attributed to helium enhanced intergranular creep crack growth, which reduces the ductility and strength, but does not affect the creep rate. The variation of tensile and creep properties is large and independent of irradiation. The minimum derived creep strength in irradiated condition drops below the values expected in the ASME Code and VdTuV Blatt. In design and safety analyses the irradiation effect on creep properties must be accounted for with an appropriate reduction factor. The predictions given, have to be verified with long-term creep tests and parts of the SNR surveillance programme. 172 figs.; 17 refs.; 58 tables

  7. Reducing the moisture effect on the creep deformation of wood by an irradiation-induced polymer impregnation method

    International Nuclear Information System (INIS)

    Chia, L.H.L.; Boey, F.Y.C.; Teoh, S.H.

    1988-01-01

    This paper reports an attempt to reduce the sensitivity of creep deformation to moisture adsorption by impregnating a tropical wood with methyl-methacrylate and subsequent polymerization by γ-irradiation. Beam specimens both of untreated wood and polymer impregnated wood were subjected to a three-point bend creep test under a constant load of 300 N at 23 ± 1 0 C, at three different humidity levels of 50 ± 5, 65 ± 5 and 85 ± 5%. A Norton-Bailey (power law) mathematical model successfully described the creep behaviour, with the creep components determined by a non-linear regression analysis. A significant reduction in the sensitivity of creep deformation to the humidity level was attained for the polymer impregnated wood. This could be explained by the ability of the impregnated polymer to form a strongly adhesive interface with the wood cell material, thereby acting as a physical barrier to reduce the movement of water to and from the wood cell material. (author)

  8. Comparison of Material Behavior of Matrix Graphite for HTGR Fuel Elements upon Irradiation: A literature Survey

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young-Woo; Yeo, Seunghwan; Cho, Moon Sung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The fuel elements for the HTGRs (i.e., spherical fuel element in pebble-bed type core design and fuel compact in prismatic core design) consists of coated fuel particles dispersed and bonded in a closely packed array within a carbonaceous matrix. This matrix is generally made by mixing fully graphitized natural and needle- or pitchcoke originated powders admixed with a binder material (pitch or phenolic resin), The resulting resinated graphite powder mixture, when compacted, may influence a number of material properties as well as its behavior under neutron irradiation during reactor operation. In the fabrication routes of these two different fuel element forms, different consolidation methods are employed; a quasi-isostatic pressing method is generally adopted to make pebbles while fuel compacts are fabricated by uni-axial pressing mode. The result showed that the hardness values obtained from the two directions showed an anisotropic behavior: The values obtained from the perpendicular section showed much higher micro hardness (176.6±10.5MPa in average) than from the parallel section ((125.6±MPa in average). This anisotropic behavior was concluded to be related to the microstructure of the matrix graphite. This may imply that the uni-axial pressing method to make compacts influence the microstructure of the matrix and hence the material properties of the matrix graphite.

  9. Leaching of 14C and 36Cl from irradiated French graphite

    International Nuclear Information System (INIS)

    Gray, W.J.; Morgan, W.C.

    1989-08-01

    The leach rates of 14 C and 36 Cl were measured on solid cylindrical samples prepared from irradiated graphite blocks supplied by the French Commissariat a l'Energie Atomique (CEA). Static leach tests were conducted in deionized water at 20 degree C for 13 weeks. The graphite samples were completely submerged in the water, and the entire volume of water was changed and analyzed at weekly intervals for the first three weeks and biweekly thereafter. Large differences in the leach rates of both 14 C and 36 Cl were observed between samples machined from the different blocks. In general, the leach rates were much higher than those measured in an earlier study with graphite obtained from a block removed from one of the Hanford reactors. The data from this study are compared with those from the previous study using the Hanford-reactor graphite. Implications of the data from both studies regarding possible rate-limiting mechanisms are discussed. 4 refs., 8 figs., 3 tabs

  10. Processing of Irradiated Graphite to Meet Acceptance Criteria for Waste Disposal. Results of a Coordinated Research Project. Companion CD-ROM

    International Nuclear Information System (INIS)

    2016-05-01

    Graphite is widely used in the nuclear industry and in research facilities and this has led to increasing amounts of irradiated graphite residing in temporary storage facilities pending disposal. This publication arises from a coordinated research project (CRP) on the processing of irradiated graphite to meet acceptance criteria for waste disposal. It presents the findings of the CRP, the general conclusions and recommendations. The topics covered include, graphite management issues, characterization of irradiated graphite, processing and treatment, immobilization and disposal. Included on the attached CD-ROM are formal reports from the participants

  11. Simultaneous heating and compression of irradiated graphite during synchrotron microtomographic imaging

    Science.gov (United States)

    Bodey, A. J.; Mileeva, Z.; Lowe, T.; Williamson-Brown, E.; Eastwood, D. S.; Simpson, C.; Titarenko, V.; Jones, A. N.; Rau, C.; Mummery, P. M.

    2017-06-01

    Nuclear graphite is used as a neutron moderator in fission power stations. To investigate the microstructural changes that occur during such use, it has been studied for the first time by X-ray microtomography with in situ heating and compression. This experiment was the first to involve simultaneous heating and mechanical loading of radioactive samples at Diamond Light Source, and represented the first study of radioactive materials at the Diamond-Manchester Imaging Branchline I13-2. Engineering methods and safety protocols were developed to ensure the safe containment of irradiated graphite as it was simultaneously compressed to 450N in a Deben 10kN Open-Frame Rig and heated to 300°C with dual focused infrared lamps. Central to safe containment was a double containment vessel which prevented escape of airborne particulates while enabling compression via a moveable ram and the transmission of infrared light to the sample. Temperature measurements were made in situ via thermocouple readout. During heating and compression, samples were simultaneously rotated and imaged with polychromatic X-rays. The resulting microtomograms are being studied via digital volume correlation to provide insights into how thermal expansion coefficients and microstructure are affected by irradiation history, load and heat. Such information will be key to improving the accuracy of graphite degradation models which inform safety margins at power stations.

  12. Deuterium migration in nuclear graphite: consequences for the behavior of tritium in Gas Cooled Reactors and for the decontamination of irradiated graphite waste

    International Nuclear Information System (INIS)

    Le-Guillou, Mael

    2014-01-01

    In France, 23 000 t of irradiated graphite that will be generated by the decommissioning of the first generation Uranium Naturel-Graphite-Gaz (UNGG) nuclear reactors are waiting for a long term management solution. This work focuses on the behavior of tritium, which is one of the main contributors to the radiological inventory of graphite waste after reactor shutdown. In order to anticipate tritium release during dismantling or waste management, it is mandatory to collect data on its migration, location and inventory. Our study is based on the simulation of tritium by implantation of approximately 3 at. % of deuterium up to around 3 μm in a virgin nuclear graphite. This material was then annealed up to 300 h and 1300 C in inert atmosphere, UNGG coolant gas and humid gas, aiming to reproduce thermal conditions close to those encountered in reactor and during waste management operations. The deuterium profiles and spatial distribution were analyzed using the nuclear reaction 2 H( 3 He,p) 4 He. The main results evidence a thermal release of implanted deuterium occurring essentially through three regimes controlled by the detrapping of atomic deuterium located in superficial or interstitial sites. The extrapolation of our data to tritium suggests that its purely thermal release during reactor operations may have been lower than 30 % and would be located close to the graphite free surfaces. Consequently, most of the tritium inventory after reactor shutdown could be trapped deeply within the irradiated graphite structure. Decontamination of graphite waste should then require temperatures higher than 1300 C, and would be more efficient in dry inert gas than in humid gas. (author)

  13. Irradiation creep of the martensitic steel no. 1.4914 between 400 deg C and 600 deg C (Mol 5B)

    International Nuclear Information System (INIS)

    Herschbach, K.; Doser, W.

    1983-01-01

    The irradiation induced creep of the martensitic steel DIN No. 1.4914 was investigated in the temperature range from 400 to 600 deg C for stresses up to 200 Mpa using the Mol 5B irradiation rig. The results point to a behavior quite different from that observed in the austenitic steels as will be discussed in detail. The creep is thermally activated and non-linearly dependent upon the applied stress. (author)

  14. Time of flight measurements of unirradiated and irradiated nuclear graphite under cyclic compressive load

    Energy Technology Data Exchange (ETDEWEB)

    Bodel, W., E-mail: william.bodel@hotmail.com [Nuclear Graphite Research Group, The University of Manchester (United Kingdom); Atkin, C. [Health and Safety Laboratory, Buxton (United Kingdom); Marsden, B.J. [Nuclear Graphite Research Group, The University of Manchester (United Kingdom)

    2017-04-15

    The time-of-flight technique has been used to investigate the stiffness of nuclear graphite with respect to the grade and grain direction. A loading rig was developed to collect time-of-flight measurements during cycled compressive loading up to 80% of the material's compressive strength and subsequent unloading of specimens along the axis of the applied stress. The transmission velocity (related to Young's modulus), decreased with increasing applied stress; and depending on the graphite grade and orientation, the modulus then increased, decreased or remained constant upon unloading. These tests were repeated while observing the microstructure during the load/unload cycles. Initial decreases in transmission velocity with compressive load are attributed to microcrack formation within filler and binder phases. Three distinct types of behaviour occur on unloading, depending on the grade, irradiation, and loading direction. These different behaviours can be explained in terms of the material microstructure observed from the microscopy performed during loading.

  15. Status of IAEA international data base on irradiated graphite properties with respect to HTR engineering issues

    International Nuclear Information System (INIS)

    Hacker, P.J.; Haag, G.

    2002-01-01

    The International Database on Irradiated Nuclear Graphite Properties contains data on the physical, chemical, mechanical and other relevant properties of graphites. Its purpose is to provide a platform that makes these properties accessible to approved users in the fields of nuclear power, nuclear safety and other nuclear science and technology applications. The database is constructed using Microsoft Access 97 software and has a controlled distribution by CD ROM to approved users. This paper describes the organisation and management of the database through administrative arrangements approved by the IAEA. It also outlines the operation of the database. The paper concludes with some remarks upon and illustrations of the usefulness of the database for the design and operation of HTR. (authors)

  16. Graphite moderator annealing of the experimental reactor for irradiation (0.5 MW)

    International Nuclear Information System (INIS)

    Oliveira Avila, Carlos Alberto de; Pires, Luis Fernando Goncalves

    1995-01-01

    This work describes an operational procedure for the annealing of the graphite moderator in the 0,5 MW Experimental Reactor for Irradiation. A theoretical methodology has been developed for calculating the temperature field during the annealing process. The equations for mass, momentum, and energy conservation for the coolant as well as for the energy conservation in the moderator are solved numerically. The energy stored in the graphite and released in the annealing is accounted for by the use of a modified source term in the energy conservation equation for the moderator. A good agreement has been found for comparisons of the calculations with annealing data from the BEPO reactor. The major parameters affecting annealing have also been determined. (author). 8 refs, 11 figs

  17. Monovacancy paramagnetism in neutron-irradiated graphite probed by 13C NMR.

    Science.gov (United States)

    Zhang, Z T; Xu, C; Dmytriieva, D; Molatta, S; Wosnitza, J; Wang, Y T; Helm, M; Zhou, Shengqiang; Kühne, H

    2017-10-20

    We report on the magnetic properties of monovacancy defects in neutron-irradiated graphite, probed by 13 C nuclear magnetic resonance spectroscopy. The bulk paramagnetism of the defect moments is revealed by the temperature dependence of the NMR frequency shift and spectral linewidth, both of which follow a Curie behavior, in agreement with measurements of the macroscopic magnetization. Compared to pristine graphite, the fluctuating hyperfine fields generated by the defect moments lead to an enhancement of the 13 C nuclear spin-lattice relaxation rate [Formula: see text] by about two orders of magnitude. With an applied magnetic field of 7.1 T, the temperature dependence of [Formula: see text] below about 10 K can well be described by a thermally activated form, [Formula: see text], yielding a singular Zeeman energy of ([Formula: see text]) meV, in excellent agreement with the sole presence of polarized, non-interacting defect moments.

  18. Method to Assess the Radionuclide Inventory of Irradiated Graphite from Gas-Cooled Reactors - 13072

    Energy Technology Data Exchange (ETDEWEB)

    Poncet, Bernard [EDF-CIDEN, 154 Avenue Thiers, CS 60018, F-69458 LYON cedex 06 (France)

    2013-07-01

    About 17,000 t of irradiated graphite waste will be produced from the decommissioning of the six French gas-cooled nuclear reactors. Determining the radionuclide (RN) content of this waste is of relevant importance for safety reasons and in order to determine the best way to manage them. For many reasons the impurity content that gave rise to the RNs in irradiated graphite by neutron activation during operation is not always well known and sometimes actually unknown. So, assessing the RN content by the use of traditional calculation activation, starting from assumed impurity content, leads to a false assessment. Moreover, radiochemical measurements exhibit very wide discrepancies especially on RN corresponding to precursor at the trace level such as natural chlorine corresponding to chlorine 36. This wide discrepancy is unavoidable and is due to very simple reasons. The level of impurity is very low because the uranium fuel used at that very moment was not enriched, so it was a necessity to have very pure nuclear grade graphite and the very low size of radiochemical sample is a simple technical constraint because device size used to get mineralization product for measurement purpose is limited. The assessment of a radionuclide inventory only based on few number of radiochemical measurements lead in most cases, to a gross over or under-estimation that is detrimental for graphite waste management. A method using an identification calculation-measurement process is proposed in order to assess a radiological inventory for disposal sizing purpose as precise as possible while guaranteeing its upper character. This method present a closer approach to the reality of the main phenomenon at the origin of RNs in a reactor, while also incorporating the secondary effects that can alter this result such as RN (or its precursor) release during reactor operation. (authors)

  19. Path dependent models to predict property changes in graphite irradiated at changing irradiation temperatures

    CSIR Research Space (South Africa)

    Kok, S

    2010-10-01

    Full Text Available Property changes occur in materials subjected to irradiation. The bulk of experimental data and associated empirical models are for isothermal irradiation. The form that these isothermal models take is usually closed form expressions in terms...

  20. Synthesis of metal free ultrathin graphitic carbon nitride sheet for photocatalytic dye degradation of Rhodamine B under visible light irradiation

    Science.gov (United States)

    Rahman, Shakeelur; Momin, Bilal; Higgins M., W.; Annapure, Uday S.; Jha, Neetu

    2018-04-01

    In recent times, low cost and metal free photocatalyts driven under visible light have attracted a lot of interest. One such photo catalyst researched extensively is bulk graphitic carbon nitride sheets. But the low surface area and weak mobility of photo generated electrons limits its photocatalytic performance in the visible light spectrum. Here we present the facile synthesis of ultrathin graphitic carbon nitride using a cost effective melamine precursor and its application in highly efficient photocatalytic dye degradation of Rhodamine B molecules. Compared to bulk graphitic carbon nitride, the synthesized ultrathin graphitic carbon nitride shows an increase in surface area, a a decrease in optical band gap and effective photogenerated charge separation which facilitates the harvest of visible light irradiation. Due to these optimal properties of ultrathin graphitic carbon nitride, it shows excellent photocatalytic activity with photocatalytic degradation of about 95% rhodamine B molecules in 1 hour.

  1. Ion irradiation of {sup 37}Cl implanted nuclear graphite: Effect of the energy deposition on the chlorine behavior and consequences for the mobility of {sup 36}Cl in irradiated graphite

    Energy Technology Data Exchange (ETDEWEB)

    Toulhoat, N., E-mail: nelly.toulhoat@univ-lyon1.fr [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); CEA/DEN, Centre de Saclay (France); Moncoffre, N. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); Bérerd, N.; Pipon, Y. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); Université de Lyon, Université Lyon, IUT Lyon-1 département chimie (France); Blondel, A. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); Andra, Châtenay-Malabry (France); Galy, N. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); Sainsot, P. [Université de Lyon, Université Lyon 1, LaMCoS, INSA-Lyon, CNRS UMR5259 (France); Rouzaud, J.-N.; Deldicque, D. [Laboratoire de Géologie de l’Ecole Normale Supérieure (ENS), Paris, UMR CNRS-ENS 8538 (France)

    2015-09-15

    Graphite is used in many types of nuclear reactors due to its ability to slow down fast neutrons without capturing them. Whatever the reactor design, the irradiated graphite waste management has to be faced sooner or later regarding the production of long lived or dose determining radioactive species such as {sup 14}C, {sup 3}H or {sup 36}Cl. The first carbon dioxide cooled, graphite moderated nuclear reactors resulted in a huge quantity of irradiated graphite waste for which the management needs a previous assessment of the radioactive inventory and the radionuclide’s location and speciation. As the detection limits of usual spectroscopic methods are generally not adequate to detect the low concentration levels (<1 ppm) of the radionuclides, we used an indirect approach based on the implantation of {sup 37}Cl, to simulate the presence of {sup 36}Cl. Our previous studies show that temperature is one of the main factors to be considered regarding the structural evolution of nuclear graphite and chlorine mobility during reactor operation. However, thermal release of chlorine cannot be solely responsible for the depletion of the {sup 36}Cl inventory. We propose in this paper to study the impact of irradiation and its synergetic effects with temperature on chlorine release. Indeed, the collision of the impinging neutrons with the graphite matrix carbon atoms induces mainly ballistic collisions. However, a small part of the recoil carbon atom energy is also transferred to the lattice through electronic excitation. This paper aims at elucidating the effects of the different irradiation regimes (ballistic and electronic) using ion irradiation, on the mobility of implanted {sup 37}Cl, taking into account the initial disorder level of the nuclear graphite.

  2. IAEA International Database on Irradiated Nuclear Graphite Properties. 7th meeting of the Technical Steering Committee

    International Nuclear Information System (INIS)

    Humbert, D.; Wickham, A.J.

    2005-06-01

    This report summarizes the Consultant Meeting '7th Meeting of the Technical Steering Committee for the International Database on Irradiated Nuclear Graphite Properties' held on 16-17 March 2005 at the IAEA Headquarters, Vienna, Austria. The purposes of the meeting were to review the matters and actions identified in the previous meeting, undertake a review of the current status of the database and to make recommendations for actions for the next year. The purposes of the meeting were fully met. This report contains the current status of the identified actions as well as a summary of the recommendations on enhancements to the database. (author)

  3. Summary report of consultants' meeting on IAEA international database on irradiated nuclear graphite properties

    International Nuclear Information System (INIS)

    Humbert, D.; Wickham, A.J.

    2007-06-01

    The '9th Meeting of the Technical Steering Committee for the International Database on Irradiated Nuclear Graphite Properties' was held on 26-27 March 2007 at the IAEA Headquarters, Vienna, Austria. All discussions, recommendations and actions of this Consultants' Meeting are recorded in this report. The purposes of the meeting were to review the matters and actions identified in the previous meeting, undertake a review of the current status of the database and make recommendations for actions for the next year. This report contains the current status of the identified actions as well as a summary of the recommendations on enhancements to the database. (author)

  4. IAEA International Database on Irradiated Nuclear Graphite Properties. 6th meeting of the Technical Steering Committee

    International Nuclear Information System (INIS)

    Humbert, D.; Wickham, A.J.

    2004-12-01

    This report summarizes the Consultant Meeting 6th Meeting of the Technical Steering Committee for the International Database on Irradiated Nuclear Graphite Properties' held on 16-17 September 2004 at Plas Tan-Y-Bwlch, Maentwrog, Gwynedd, UK. The purposes of the meeting were to review the matters and actions identified in the previous meeting, undertake a review of the current status of the database and to make recommendations for actions for the next year. The purposes of the meeting were fully met. This report contains the current status of the identified actions as well as a summary of the recommendations on enhancements to the database. (author)

  5. Effect of reactor irradiation on long-term strength and creep of 0Kh16N15M3B steel under plane stressed state

    International Nuclear Information System (INIS)

    Khristov, G.P.; Kosov, B.D.

    1982-01-01

    The paper deals with analysis of results of experimental studies in creep of the austenitic OKh16n15m3b steel with various size of initial-structure grain under conditions of high-intensity reactor irradiation and control tests. It is suggested to consider the material initial structure effect on intensity of minimum creep rates both under ordinary and intrareactor conditions of loading by means of the function grain size effect on the equivalent stress. It is shown that the criterial expression previously suggested by the authors is invariant to the type of stressed and structural states and relative to intensity of minimal creep rates. It is established that the creep rate of the irradiated steel may be calculated from dependence for nonirradiated steel using as an argument a certain reduced equivalent stress which is a function of the acting stress and irradiation parameter

  6. Effects of 14 MeV neutron irradiation on creep of nickel and niobium

    International Nuclear Information System (INIS)

    Barmore, W.; Ruotola, A.; Raymond, E.; Mukherjee, A.

    1983-01-01

    Flux, stress and temperature effects on the creep strength of nickel and niobium were observed in situ at the RTNS-II 14 MeV neutron source at Lawrence Livermore National Laboratory. Creep test were done on Ni and Nb near 0.3 Tsub(m) with stresses to 280 MPa in a high vacuum test unit using a digital computer for control and data acquisition. Cyclic flux tests produced dramatic changes in creep rate. This creep behavior is attributed to the point defect fluctuations in the crystal structure. Analysis of creep and stress relaxation under steady state flux indicates that an intermediate temperature, thermally activated deformation mechanism is rate controlling. (orig.)

  7. Effects of high temperature neutron irradiation on the physical, chemical and mechanical properties of fine-grained isotropic graphite

    International Nuclear Information System (INIS)

    Matsuo, H.; Nomura, S.; Imai, H.; Oku, T.; Eto, M.

    1987-01-01

    Effects of neutron irradiation on the dimensional change, coefficient of thermal expansion(CTE), thermal conductivity, corrosion rate, Young's modulus and strengths were studied for the candidate graphite material IG-110 of the experimental very high temperature gas-cooled reactor(VHTR) after irradiation at 585 - 1273 deg C to neutron fluences of up to about 3 x 10 25 n/m 2 (E > 29 fJ) in the JMTR and JRR-2, and to about 7 x 10 25 n/m 2 (E > 29 fJ) in the HFR. The results were compared with the irradiation behaviors of other graphites. Dimensional shrinkage was observed in the whole irradiation temperature range, showing lower value than 2 %. The shrinkage rate showed the minimum in the irradiation temperature of around 850 deg C, followed by the increase for the samples irradiated at higher temperatures. The dimensional stability of the material was clarified to be almost the same with that of H451 graphite. The CTE, thermal resistivity and Young's modulus increased in the early stage of irradiation and then only the CTE decreased while the thermal resistivity and Young's modulus levelled off with further irradiation. The neutron fluence showing the maximum CTE shifted to the lower fluence with increasing irradiation temperature. The increases of both thermal resistivity and Young's modulus were remarkable for the samples irradiated at lower temperatures. Compressive and bending strengths measured at room temperature increased after irradiation as well. The corrosion rate with water-vapor of 0.65 % in helium at high temperatures decreased owing to irradiation and the reduction was independent of irradiation temperature and neutron fluence. The activation energy for the reaction was estimated to be the same before and after irradiation. (author)

  8. Post Irradiation Examination Results of the NT-02 Graphite Fins NUMI Target

    Energy Technology Data Exchange (ETDEWEB)

    Ammigan, K. [Fermilab; Hurh, P. [Fermilab; Sidorov, V. [Fermilab; Zwaska, R. [Fermilab; Asner, D. M. [PNL, Richland; Casella, Casella,A.M [PNL, Richland; Edwards, D. J. [PNL, Richland; Schemer-Kohrn, A. L. [PNL, Richland; Senor, D. J. [PNL, Richland

    2017-02-10

    The NT-02 neutrino target in the NuMI beamline at Fermilab is a 95 cm long target made up of segmented graphite fins. It is the longest running NuMI target, which operated with a 120 GeV proton beam with maximum power of 340 kW, and saw an integrated total proton on target of 6.1 1020. Over the last half of its life, gradual degradation of neutrino yield was observed until the target was replaced. The probable causes for the target performance degradation are attributed to radiation damage, possibly including cracking caused by reduction in thermal shock resistance, as well as potential localized oxidation in the heated region of the target. Understanding the long-termstructural response of target materials exposed to proton irradiation is critical as future proton accelerator sources are becoming increasingly more powerful. As a result, an autopsy of the target was carried out to facilitate post-irradiation examination of selected graphite fins. Advanced microstructural imaging and surface elemental analysis techniques were used to characterize the condition of the fins in an effort to identify degradation mechanisms, and the relevant findings are presented in this paper.

  9. An investigation of the electron irradiation of graphite in a helium atmosphere using a modified electron microscope

    International Nuclear Information System (INIS)

    Burden, A.P.; Hutchison, J.L.

    1997-01-01

    The behaviour of graphite particles immersed in helium gas and irradiated with an electron-beam has been investigated. Because this treatment was performed in a modified high resolution transmission electron microscope, the rapid morphological and microstructural changes that occurred could be directly observed. The results have implications for future controlled environment microscopy of carbonaceous materials and the characterisation of such microscopes. It is also shown that the processes can provide insight into ion-irradiation induced damage of graphite and the mechanism of fullerene generation. (Author)

  10. Radiation effects on time-dependent deformation: Creep and growth

    International Nuclear Information System (INIS)

    Simonen, E.P.

    1989-03-01

    Observations of irradiation creep strain as well as irradiation growth strain and related microstructures are reviewed and compared to mechanisms for radiation effects on time-dependent deformation. Composition, microstructure, stress and temperature affect irradiation creep less than thermal creep. Irradiation creep rates can often dominate thermal creep rates, particularly at low temperatures and low stresses. Irradiation creep mechanisms are classified in two general categories: (1) stress-induced preferential absorption and (2) climb-glide. In the former, creep results from dislocation climb, whereas in the latter, creep results from dislocation glide. The effects of irradiation creep on failure modes in nuclear environments are discussed. 53 refs., 18 figs., 1 tab

  11. Technical specifications (replaces note T.62). Irradiation of graphite at ambient temperature, Note T. 76; Specification technique, (Annule et remplace la note T. 62), Irradiation de graphite a temperature ambiante, Note T. 76

    Energy Technology Data Exchange (ETDEWEB)

    Reseau, R A [Services des grandes piles experimentales, Section ' Physique et Experimentation, Saclay (France)

    1962-12-15

    The objective is to study the effects of fast neutron irradiation of different graphite samples. The irradiation conditions should be as follows: integral fast neutron flux should be higher than 10{sup 20} neutrons/cm{sup 2}, the reactor should operate at steady state for 15 days, the temperature od samples should not be higher than 100 deg C, preferably 80 deg C. Note T. 62 which is replaced by this Note is attached.

  12. Non-isothermal irradiation creep of nickel alloys Inconel 706 and PE-16

    International Nuclear Information System (INIS)

    Gilbert, E.R.; Chin, B.A.

    1984-06-01

    The results of in-reactor step temperature change experiments conducted on two nickel alloys, PE-16 and Inconel 706, were evaluated to determine the creep behavior under nonisothermal conditions. The effect of the temperature changes was found to be significantly different for the two alloys. Following a step temperature change, the creep rate of PE-16 adjusted to the rate found in isothermal tests at the new temperature. In contrast for Inconel 706, a reduction in temperature from 540 to 425 0 C produced a 300% increase in creep above that measured at 540 0 C in isothermal tests. The response of in-reactor creep in Inconel 706 to temperature changes was attributed to the dissolution of the gamma double-prime phase and subsequent loss of precipitation-strengthening at temperatures below 500 C

  13. Report on fundamental modeling of irradiation-induced swelling and creep in FeCrAl alloys

    Energy Technology Data Exchange (ETDEWEB)

    Kohnert, Aaron A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Dasgupta, Dwaipayan [Univ. of Tennessee, Knoxville, TN (United States); Wirth, Brian [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Linton, Kory D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-09-23

    In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys due to much slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. As a continuation of the development for these alloys, the material response must be demonstrated to provide suitable radiation stability, in order to ensure that there will not be significant dimensional changes (e.g., swelling), as well as quantifying the radiation hardening and radiation creep behavior. In this report, we describe the use of cluster dynamics modeling to evaluate the defect physics and damage accumulation behavior of FeCrAl alloys subjected to neutron irradiation, with a particular focus on irradiation-induced swelling and defect fluxes to dislocations that are required to model irradiation creep behavior.

  14. Thermal/hydraulic bowing stability analysis of grid-supported multi-pin bundles with differential swelling and irradiation creep

    International Nuclear Information System (INIS)

    McAreavey, G.

    1977-01-01

    Azimuthal variations of clad temperature in fuel pin bundles leads to pin bowing by differential thermal expansion. During irradiation in a fast flux further possibly more severe bowing is caused by differential neutron induced voidage swelling, which, being temperature sensitive, will also vary azimuthally. The problem of pin bowing in a fuel element cluster involves consideration of the thermal/hydraulic behaviour, allowing for both inherent and induced clad temperature non-uniformities, coupled with the restrained bowing behaviour, including differential thermal expansion, differential swelling, and irradiation creep. All pins must be considered simultaneously. In the temperature and stress ranges of interest thermal creep may be neglected. An existing computer code, IAMBIC solves the zero time thermal bowing problem for a cluster of up to 61 pins on hexagonal pitch, with up to 21 supports at arbitrary axial spacing. The present paper describes the basis of TRIAMBIC, a time dependent code which analyses the irradiation induced effects in fuel pin bunbles due to fast neutrons. (Auth.)

  15. Impact of creep-fatigue interaction on the lifetime of a dispersion strengthened copper alloy in unirradiated and irradiated conditions

    International Nuclear Information System (INIS)

    Singh, B.N.; Toft, P.; Stubbins, J.F.

    2001-06-01

    Creep-fatigue interaction behaviour of a dispersion strengthened copper alloy was investigated at 22 and 250 deg. C. To determine the effect of irradiation a number of fatigue specimens were irradiated at 250 deg. C to a dose level of 0.3 dpa and were tested at 250 deg. C. The creep-fatigue interaction was simulated by applying a certain hold-time on both tension and compression sides of the cyclic loading with a frequency of 0.5 Hz. Hold-times of 0,2, 5, 10, 100 and 1000 seconds were used. For a given hold-time, the real lifetime and the number of cycles to failure were determined at different strain amplitudes. Post-deformation micro-structures and fracture surfaces were investigated using transmission and scanning electron microscopes, respectively. The main results of these investigations are presented and their implications are briefly discussed in the present report. The central conclusion emerging from the present work is that a hold-time of 10 seconds or less causes a drastic decrease in the real lifetime as well as in the number of cycles to failure, particularly at low levels of strain amplitudes. A combination of higher temperature, higher strain amplitude and longer hold-time, on the other hand, may lead to an improvement in the lifetime. The irradiation at 250 deg. C to a dose level of 0.3 dpa does not play any significant role in determining the lifetime under creep-fatigue testing conditions. (au)

  16. Release of fission products and post-pile creep behaviour of irradiated fuel rods stored under dry conditions

    International Nuclear Information System (INIS)

    Kaspar, G.; Peehs, M.; Bokelmann, R.; Jorde, D.; Schoenfeld, H.; Haas, W.; Bleier, A.; Rutsch, F.

    1985-06-01

    The release of moisture and fission products (Kr-85, H-3 and I-129) under dry storage conditions has been examined on six fuel rods which have become defective in the reactor. During the examinations, inert conditions prevailed and limited air inlet was allowed temporarily. The storage temperature was 400 0 C. The residual moisture content of the fuel rods was approx. 5 g. At the beginning of the test, the total moisture content and 0,05% (max.) of the fission gas inventory were released. Under inert conditions, fission gas was not released during a prolonged period of time. Under oxidizing conditions, however, fission gas was released in the course of UO 2 oxidation. Post-pile creep of Zircaloy cladding tubes was measured at temperatures between 350 and 395 0 C and interval gauge pressures between 69 and 110 bar. The creep curves indicate that the irradiated cladding tube specimens still bear internal residual stresses which contribute through their relaxation to the post-pile creep. (orig.) [de

  17. Distribution of 60Co and 54Mn in graphite material of irradiated HTGR fuel assemblies

    International Nuclear Information System (INIS)

    Hayashi, Kimio; Kikuchi, Teruo; Kobayashi, Fumiaki; Minato, Kazuo; Fukuda, Kousaku; Ikawa, Katsuichi; Iwamoto, Kazumi

    1984-05-01

    Distribution of 60 Co and 54 Mn was measured in the graphite sleeves and blocks of the third and fourth HTGR fuel assemblies irradiated in the Oarai Gas Loop-1 (OGL-1), which is a high temperature inpile gas loop installed in the Japan Materials Testing Reactor (JMTR) of Japan Atomic Energy Research Institute (JAERI). Axial and circumferential profiles were obtained by gamma spectrometry, and radial profiles by lathe sectioning with gamma spectrometry. Distribution of 60 Co is in good agreement with that of thermal neutron flux, and the Co content in the graphite is estimated to be -- 1 x 10 -9 in weight fraction. Concentration of 54 Mn decreases toward the axial center in its axial profile, and radially is almost uniform inside and appreciably higher at free surfaces. An estimated Fe content of --10 -8 in wight fraction is smaller by two orders of magnitude than that from chemical analysis. Higher concentraion of 60 Co and 54 Mn at the free surfaces suggests the importance of transportation process of these nuclides in the coolant loop. (author)

  18. Cyclic fatigue of near-isotopic graphite: influence of stress cycle and neutron irradiation

    International Nuclear Information System (INIS)

    Price, R.J.

    1977-11-01

    Near-isotropic graphites H-451 and PGX were tested in uniaxial cyclic fatigue, and fatigue life (S-N) curves were generated to a maximum of 10 5 cycles. The stress ratio, R (minimum stress during a cycle divided by maximum stress) ranged from -1 to +0.5. With R = - 1, the homologous stress limits (maximum applied fatigue stress divided by the tensile strength) for 50% specimen survival to 10 5 cycles averaged 0.63 in the axial direction and 0.74 in the radial direction. Corresponding homologous stress limits for 99% specimen survival (99/95 tolerance limits) were 0.48 and 0.53. Higher R-values resulted in longer fatigue lives and increased stress limits. H-451 graphite specimens irradiated with fast neutrons at 1173 to 1263 0 K at fluences of up to 10 26 n/m 2 (equivalent fission fluence) showed fatigue stress limits of about twice the unirradiated levels when the unirradiated tensile strength was used as the basis for normalization

  19. Monovacancy paramagnetism in neutron-irradiated graphite probed by 13C NMR.

    Science.gov (United States)

    Zhang, Zhi Tao; Xu, C; Dmytriieva, Daryna; Molatta, Sebastian; Wosnitza, J; Wang, Y T; Helm, Manfred; Zhou, Shengqiang; Kuehne, Hannes

    2017-09-18

    We report on the magnetic properties of monovacancy defects in neutron-irradiated graphite, probed by $^{13}$C nuclear magnetic resonance spectroscopy. The bulk paramagnetism of the defect moments is revealed by the temperature dependence of the NMR frequency shift and spectral linewidth, both of which follow a Curie behavior, in agreement with measurements of the macroscopic magnetization. Compared to pristine graphite, the fluctuating hyperfine fields generated by the defect moments lead to an enhancement of the $^{13}$C nuclear spin-lattice relaxation rate $1/T_{1}$ by about two orders of magnitude. With an applied magnetic field of 7.1 T, the temperature dependence of $1/T_{1}$ below about 10 K can well be described by a thermally activated form, $1/T_{1}\\propto\\exp(-\\Delta/k_{B}T)$, yielding a singular Zeeman energy of ($0.41\\pm0.01$) meV, in excellent agreement with the sole presence of polarized, non-interacting defect moments. © 2017 IOP Publishing Ltd.

  20. Surface structure modification of single crystal graphite after slow, highly charged ion irradiation

    Science.gov (United States)

    Alzaher, I.; Akcöltekin, S.; Ban-d'Etat, B.; Manil, B.; Dey, K. R.; Been, T.; Boduch, P.; Rothard, H.; Schleberger, M.; Lebius, H.

    2018-04-01

    Single crystal graphite was irradiated by slow, highly charged ions. The modification of the surface structure was studied by means of Low-Energy Electron Diffraction. The observed damage cross section increases with the potential energy, i.e. the charge state of the incident ion, at a constant kinetic energy. The potential energy is more efficient for the damage production than the kinetic energy by more than a factor of twenty. Comparison with earlier results hints to a strong link between early electron creation and later target atom rearrangement. With increasing ion fluence, the initially large-scale single crystal is first transformed into μ m-sized crystals, before complete amorphisation takes place.

  1. Models of bending strength for Gilsocarbon graphites irradiated in inert and oxidising environments

    International Nuclear Information System (INIS)

    Eason, Ernest D.; Hall, Graham N.; Marsden, Barry J.; Heys, Graham B.

    2013-01-01

    This paper presents the development and validation of an empirical model of fast neutron damage and radiolytic oxidation effects on bending strength for the moulded Gilsocarbon graphites used in Advanced Gas-cooled Reactors (AGRs). The inert environment model is based on evidence of essentially constant strength as fast neutron dose increases in inert environment. The model of combined irradiation and oxidation calibrates that constant along with an exponential function representing the degree of radiolytic oxidation as measured by weight loss. The change in strength with exposure was found to vary from one AGR station to another. The model was calibrated to data on material trepanned from AGR moderator bricks after varying operating times

  2. The irradiation behaviour of boron carbide/graphite between 800 and 1,1000C

    International Nuclear Information System (INIS)

    Hattenbach, K.; Hilgendorff, W.; Weiler, K.; Zimmermann, H.U.

    1975-01-01

    64 samples of boron carbide/graphite, a material used as burnable poison in high temperature reactors, were irradiated at temperatures between 800 and 1,100 0 C up to a fluence of 1-2 x 10 20 nvt. The following post-investigations were extended to dimensional measurements to determime a possible swelling or shrinking of the pellet, corrosion tests in completely desalinated water at 300 0 C, preparation of metallographic microsections to check for crack formation, determination of the helium hold back power and the thus involved gas chromatic analysis, as well as burn-up determinations by determining the boron 10/boron 11 ratio and the lithium concentration. (orig./LN) [de

  3. Graphite behaviour in relation to the fuel element design

    Energy Technology Data Exchange (ETDEWEB)

    Everett, M. R. [OECD High Temperature Reactor Project Dragon, Winfrith (United Kingdom); Manzel, R. [OECD High Temperature Reactor Project Dragon, Winfrith (United Kingdom); Blackstone, R. [Reactor Centrum, Petten (Netherlands); Delle, W. [Kernforschungsanlage, Juelich (Germany); Lungagnani, V. [Joint Nuclear Research Centre, Euratom, Petten (Netherlands); Krefeld, R. [Joint Nuclear Research Centre, Euratom, Petten (Netherlands)

    1969-09-01

    The first designs of H.T.R. power reactors will probably use a Gilsocarbon based graphite for both the moderator/carrier blocks and for the fuel tubes. The initial physical properties and changes of dimensions, thermal expansion coefficient, Young*s modulus, and thermal conductivity on irradiation of Gilsocarbon graphites to typical reactor dwell-time fast neutron doses of 4 * 1021 cm -2 Ni dose Dido equivalent are given and values for the irradiation creep constant are presented. The influence of these property changes and those of chemical corrosion are considered briefly in relation to the present fuel element designs. The selection of an eventual less costly replacement graphite for Gilsocarbon graphite is discussed in terms of materials properties.

  4. Behavior of LASL-made graphite, ZrC, and ZrC-containing coated particles in irradiation tests HT-28 and HT-29

    International Nuclear Information System (INIS)

    Reiswig, R.D.; Wagner, P.; Hollabaugh, C.M.; White, R.W.; O'Rourke, J.A.; Davidson, K.V.; Schell, D.H.

    1976-01-01

    Three types of materials, extruded graphite, hot-pressed ZrC, and particles with ZrC coatings, were irradiated in ORNL High Fluence Isotope Reactor Irradiation tests HT-28 and HT-29. The ZrC seemed unaffected. The graphite changed in dimensions, x-ray diffraction parameters, and thermal conductivity. The four types of coated particles tested all resisted the irradiation well, except one set of particles with double-graded C-ZrC-C coats. Overall, the results were considered encouraging for use of ZrC and extruded graphite fuel matrices. 16 fig

  5. Effects of fuel particle size and fission-fragment-enhanced irradiation creep on the in-pile behavior in CERCER composite pellets

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Yunmei [Institute of Mechanics and Computational Engineering, Department of Aeronautics and Astronautics, Fudan University, Shanghai 200433 (China); Ding, Shurong, E-mail: dsr1971@163.com [Institute of Mechanics and Computational Engineering, Department of Aeronautics and Astronautics, Fudan University, Shanghai 200433 (China); Zhang, Xunchao; Wang, Canglong; Yang, Lei [Institute of Modern Physics, Chinese Academy of Sciences, Lanzhou 730000 (China)

    2016-12-15

    The micro-scale finite element models for CERCER pellets with different-sized fuel particles are developed. With consideration of a grain-scale mechanistic irradiation swelling model in the fuel particles and the irradiation creep in the matrix, numerical simulations are performed to explore the effects of the particle size and the fission-fragment-enhanced irradiation creep on the thermo-mechanical behavior of CERCER pellets. The enhanced irradiation creep effect is applied in the 10 μm-thick fission fragment damage matrix layer surrounding the fuel particles. The obtained results indicate that (1) lower maximum temperature occurs in the cases with smaller-sized particles, and the effects of particle size on the mechanical behavior in pellets are intricate; (2) the first principal stress and radial axial stress remain compressive in the fission fragment damage layer at higher burnup, thus the mechanism of radial cracking found in the experiment can be better explained. - Highlights: • A grain-scale gas swelling model considering the development of recrystallization and resolution is adopted for particles. • The influence of fission-gas-induced porosity is considered in the constitutive relations for particles. • A simulation method is developed for the multi-scale thermo-mechanical behavior. • The effects of fuel particle size and fission-fragment-enhanced irradiation creep are investigated in pellets.

  6. Experimental approach and micro-mechanical modeling of the creep behavior of irradiated zirconium alloys; Approche experimentale et modelisation micromecanique du comportement en fluage des alliages de zircomium irradies

    Energy Technology Data Exchange (ETDEWEB)

    Ribis, J

    2007-12-15

    The fuel rod cladding, strongly affected by microstructural changes due to irradiation such as high density of dislocation loops, is strained by the end-of-life fuel rod internal pressure and the potential release of fission gases and helium during dry storage. Within the temperature range that is expected during dry interim storage, cladding undergoes long term creep under over-pressure. So, in order to have a predictive approach of the behavior of zirconium alloys cladding in dry storage conditions it is essential to take into account: initial dislocation loops, thermal annealing of loops and creep straining due to over pressure. Specific experiments and modelling for irradiated samples have been developed to improve our knowledge in that field. A Zr-1%Nb-O alloy was studied using fine microstructural investigations and mechanical testing. The observations conducted by transmission electron microscopy show that the high density of loops disappears during a heat treatment. The loop size becomes higher and higher while their density falls. The microhardness tests reveal that the fall of loop density leads to the softening of the irradiated material. During a creep test, both temperature and applied stress are responsible of the disappearance of loops. The loops could be swept by the activation of the basal slip system while the prism slip system is inhibited. Once deprived of loops, the creep properties of the irradiated materials are closed to the non irradiated state, a result whose consequence is a sudden acceleration of the creep rate. Finally, a micro-mechanical modeling based on microscopic deformation mechanisms taking into account experimental dislocation loop analyses and creep test, was used for a predictive approach by constructing a deformation mechanism map of the creep behavior of the irradiated material. (author)

  7. Ion irradiation used as surrogate of neutron irradiation in graphite: Consequences on 14C and 36Cl behavior and structural evolution

    Science.gov (United States)

    Galy, N.; Toulhoat, N.; Moncoffre, N.; Pipon, Y.; Bérerd, N.; Ammar, M. R.; Simon, P.; Deldicque, D.; Sainsot, P.

    2018-04-01

    Graphite has been widely used as neutron moderator, reflector or fuel matrix in different types of reactors such as gas cooled nuclear reactors (UNGG, Magnox, AGR), RBMK reactors or high temperature gas cooled reactors. Their operation produces a great quantity of irradiated graphite or other carbonaceous waste (around 250,000 tons worldwide) that requires a special management strategy. In the case of disposal, which is a current management strategy, two main radionuclides, 14C and 36Cl might be dose determining at the outlet. Particular attention is paid to 14C due to its long half-life (T∼5730 years) [1] and as major contributor to the radioactive dose. 14C has two main production routes, i) transmutation of nitrogen (14N(n,p)14C) where nitrogen is mainly adsorbed at the surfaces of the irradiated graphite; ii) activation of carbon from the matrix (13C(n,γ)14C). According to leaching tests, it was shown that even if the quantity of 14C released in the solution is low (less than 1% of the initial inventory), around 30% is in the organic form that would be mobile in repository conditions [2,3]. 36Cl is mainly produced through the activation of 35Cl (35Cl(n,γ)36Cl) which is an impurity in nuclear graphite. Its activity is low but it might be highly mobile in clay host rocks. Thus, in order to make informed decisions about the best management process and to anticipate potential radionuclide dissemination during dismantling and in the repository, it is necessary to collect information on 14C and 36Cl location and speciation in graphite, after reactor closure. The goal of the present paper is therefore to use ion irradiation to simulate neutron irradiation and to evaluate the irradiation effects on the behavior of 36Cl and 14C as well as on the induced graphite structure modifications. For that, to understand and model the underlying mechanisms, we used an indirect approach based on 13C or 37Cl implantation to simulate the respective presence of 14C or 36Cl. These

  8. Deuterium migration in nuclear graphite: Consequences for the behavior of tritium in CO{sub 2}-cooled reactors and for the decontamination of irradiated graphite waste

    Energy Technology Data Exchange (ETDEWEB)

    Le Guillou, M. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); Agence nationale pour la gestion des déchets radioactifs, DRD/CM – 1-7, rue Jean Monnet, Parc de la Croix-Blanche, F-92298 Châtenay-Malabry cedex (France); Toulhoat, N., E-mail: nelly.toulhoat@univ-lyon1.fr [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); CEA/DEN – Centre de Saclay, F-91191 Gif-sur-Yvette cedex (France); Pipon, Y. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); Institut Universitaire Technologique, Université Claude Bernard Lyon 1, Université de Lyon – 43, boulevard du 11 novembre 1918, F-69622 Villeurbanne cedex (France); Moncoffre, N. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); Khodja, H. [Laboratoire d’Etude des Eléments Légers, CEA/DSM/IRAMIS/NIMBE, UMR 3299 SIS2M – Centre de Saclay, F-91191 Gif-sur-Yvette cedex (France)

    2015-06-15

    In this paper, we aim at understanding tritium behavior in the graphite moderator of French CO{sub 2}-cooled nuclear fission reactors (called UNGG for “Uranium Naturel-Graphite-Gaz”) to get information on its distribution and inventory in the irradiated graphite waste after their dismantling. These findings should be useful both to improve waste treatment processes and to foresee tritium behavior during reactor decommissioning and waste disposal operations. The purpose of the present work is to elucidate the effects of temperature on the behavior of tritium during reactor operation. Furthermore, it aims at exploring options of thermal decontamination. For both purposes, annealing experiments were carried out in inert atmosphere as well as in thermal conditions as close as possible to those encountered in UNGG reactors and in view of a potential decontamination in humid gas. D{sup +} ions were implanted into virgin nuclear graphite in order to simulate tritium displaced from its original structural site through recoil during reactor operation. The effect of thermal treatments on the mobility of the implanted deuterium was then investigated at temperatures ranging from 200 to 1200 °C, in inert atmosphere (vacuum or argon), in a gas simulating the UNGG coolant gas (mainly CO{sub 2}) or in humid nitrogen. Deuterium was analyzed by Nuclear Reaction Analysis (NRA) both at millimetric and micrometric scales. We have identified three main stages for the deuterium release. The first one corresponds to deuterium permeation through graphite open pores. The second and third ones are controlled by the progressive detrapping of deuterium located at different trapping sites and its successive migration through the crystallites and along crystallites and coke grains edges. Extrapolating the thermal behavior of deuterium to tritium, the results show that the release becomes significant above the maximum UNGG reactor temperature of 500 °C and should be lower than 30% of the

  9. Temperature and radiolytic corrosion effects on the chlorine behaviour in nuclear graphite: consequences for the disposable of irradiated graphite from UNGG reactors

    International Nuclear Information System (INIS)

    Vaudey, C.E.

    2010-10-01

    This work concerns the dismantling of the UNGG reactor which have produced around 23 000 t of graphite wastes that ave to be disposed of according to the French law of June 206. These wastes contain two long-lived radionuclides ( 14 C and 36 Cl) which are the main long term dose contributors. In order to get information about their inventory and their long term behaviour in case of water ingress into the repository, it is necessary to determine their location and speciation in the irradiated graphite after the reactor shutdown. This work concerns the study of 36 Cl. The main objective is to reproduce its behaviour during reactor operation. For that purpose, we have studied the effects of temperature and radiolytic corrosion independently. Our results show a rapid release of around 20% 36 Cl during the first hours of reactor operation whereas a much slower release occurs afterwards. We have put in evidence two types of chlorine corresponding to two different chemical forms (of different thermal stabilities) or to two locations (of different accessibilities). We have also shown that the radiolytic corrosion seems to enhance chlorine release, whatever the irradiation dose. Moreover, the major chemical form of chlorine is inorganic. (author)

  10. HTGR Fuels and Core Development Program. Quarterly progress report for the period ending August 31, 1977. [Graphite and fuel irradiation; fission product release

    Energy Technology Data Exchange (ETDEWEB)

    1977-09-01

    The work reported includes studies of reactions between core materials and coolant impurities, basic fission product transport mechanisms, core graphite development and testing, the development and testing of recyclable fuel systems, and physics and fuel management studies. Materials studies include irradiation capsule tests of both fuel and graphite. Experimental procedures and results are discussed and data are presented.

  11. Strategy for Handling and Treatment of INPP RBMK-1500 Irradiated Graphite

    International Nuclear Information System (INIS)

    Oryšaka, A.

    2016-01-01

    There are two RBMK-1500 water-cooled graphite-moderated channel-type power reactors at Ignalina NPP. After the final shutdown of the INPP, radioactive i-graphite dismantling, handling, conditioning, storage and disposal is an important part of the decommissioning activities. The core of the INPP unit 1 and 2 contains about 3600 tons of i-graphite. Formation of activation products strongly depends on the contents of impurities, operational mode and concentration of impurities in the graphite. The case study for INPP envisages the analysis of possibilities of graphite handling and treatment in the context of immediate decommissioning. (author)

  12. Thermal desorption spectroscopy of pyrolytic graphite cleavage faces after keV deuterium irradiation at 330-1000 K

    International Nuclear Information System (INIS)

    Gotoh, Y.; Yamaki, T.; Tokiguchi, K.

    1992-01-01

    Thermal desorption spectroscopy (TDS) measurements were made on D 2 and CD 4 from surface layers of pyrolytic graphite cleavage faces after 3 keV D + 3 irradiation to 1.5 x 10 18 D/cm 2 at irradiation temperatures from 330 to 1000 K. Thermal desorption of both D 2 and CD 4 was observed to rise simultaneously at around 700 K. The D 2 peak was found at T m = 900-1000 K, while the CD 4 peak appeared at a lower temperature, 800-840 K. The T m for the D 2 TDS increased, while that for the CD 4 decreased with increasing irradiation temperature. These results obviously indicate that the D 2 desorption is detrapping/recombination limited, while the CD 4 desorption is most likely to be diffusion limited. The amount of thermally desorbed D 2 after the D + irradiation was observed to monotonously decrease as the irradiation temperature was increased from 330 to 1000 K. These tendencies agreed with previous results for the irradiation temperature dependencies of both C1s chemical shift (XPS) and the interlayer spacing, d 002 (HRTEM), on the graphite basal face. (orig.)

  13. Phonon scattering in graphite

    International Nuclear Information System (INIS)

    Wagner, P.

    1976-04-01

    Effects on graphite thermal conductivities due to controlled alterations of the graphite structure by impurity addition, porosity, and neutron irradiation are shown to be consistent with the phonon-scattering formulation 1/l = Σ/sub i equals 1/sup/n/ 1/l/sub i/. Observed temperature effects on these doped and irradiated graphites are also explained by this mechanism

  14. Influence of helium embrittlement on post-irradiation creep rupture behaviour of austenitic and martensitic stainless steels

    International Nuclear Information System (INIS)

    Wassilew, C.

    1982-01-01

    The author has investigated the influence of helium embrittlement on the creep rupture properties of the austenitic stainless steels 1.4970 and 1.4962 and the martensitic stainless steel 1.4914 after irradiation in the BR-2 reactor in Mol, Belgium. The results show that austenitic steels react much more strongly to the embrittlement effect of the helium than do martensitic steels. The causes of the lower embrittlement tendency of the martensitic than of both austenitic stainless steels were analysed carefully. A new embrittlement model was developed on the basis of data derived from the creep rupture experiments, and reinforced by a simple metallographic investigation of the fracture zone and its immediate environment. This model pays specific attention to the role of the twin planes as the most efficient area of increased vacancy production, and takes into account the ability of the twin boundaries to transport these vacancies with reduced energy and low loss into the high-angle grain boundaries. (author)

  15. Creep tests of AISI 316 stainless steel irradiated by alpha particles of 28 MeV

    International Nuclear Information System (INIS)

    Segura, E.; Lucki, G.

    1986-01-01

    He-embrittlement effect in AISI 316 SS type throught creep tests performed with annealed and cold worked thin specimens is analized. Measurements were carried out at 700 and 750 0 C, stress of 100 MPa in vacuum better than 10 -5 torr. The He-implantations were made with the cyclotron CV-28 IPEN-CNEN/SP. Using an alpha-particle beam of 28 MeV, with concentration of 26 appm. From the valves of rupture deformation, epsilon sub(R), and rupture time, t sub(R), it was verified that he had a great effect on the operational life and ductility of this material. (Author) [pt

  16. Structural performance of a graphite blanket in fusion reactors

    International Nuclear Information System (INIS)

    Wolfer, W.G.; Watson, R.D.

    1978-01-01

    Irradiation of graphite in a fusion reactor causes dimensional changes, enhanced creep, and changes in elastic properties and fracture strength. Temperature and flux gradients through the graphite blanket structure produce differential distortions and stress gradients. An inelastic stress analysis procedure is described which treats these variations of the graphite properties in a consistent manner as dictated by physical models for the radiation effects. Furthermore, the procedure follows the evolution of the stress and fracture strength distributions during the reactor operation as well as for possible shutdowns at any time. The lifetime of the graphite structure can be determined based on the failure criterion that the stress at any location exceeds one-half of the fracture strength. This procedure is applied to the most critical component of the blanket module in the SOLASE design

  17. AGC 2 Irradiated Material Properties Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Rohrbaugh, David Thomas [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-05-01

    The Advanced Reactor Technologies Graphite Research and Development Program is conducting an extensive graphite irradiation experiment to provide data for licensing of a high temperature reactor (HTR) design. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor designs. , Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphite grades have been developed and are considered suitable candidates for new HTR reactor designs. To support the design and licensing of HTR core components within a commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance are required for the physical, mechanical, and thermal properties of each graphite grade, with a specific emphasis on data accounting for the life limiting effects of irradiation creep on key physical properties of the HTR candidate graphite grades. Further details on the research and development activities and associated rationale required to qualify nuclear grade graphite for use within the HTR are documented in the graphite technology research and development plan.

  18. Modeling of irradiated graphite {sup 14}C transfer through engineered barriers of a generic geological repository in crystalline rocks

    Energy Technology Data Exchange (ETDEWEB)

    Poskas, Povilas; Grigaliuniene, Dalia, E-mail: Dalia.Grigaliuniene@lei.lt; Narkuniene, Asta; Kilda, Raimondas; Justinavicius, Darius

    2016-11-01

    There are two RBMK-1500 type graphite moderated reactors at the Ignalina nuclear power plant in Lithuania, and they are under decommissioning now. The graphite cannot be disposed of in a near surface repository, because of large amounts of {sup 14}C. Therefore, disposal of the graphite in a geological repository is a reasonable solution. This study presents evaluation of the {sup 14}C transfer by the groundwater pathway into the geosphere from the irradiated graphite in a generic geological repository in crystalline rocks and demonstration of the role of the different components of the engineered barrier system by performing local sensitivity analysis. The speciation of the released {sup 14}C into organic and inorganic compounds as well as the most recent information on {sup 14}C source term was taken into account. Two alternatives were considered in the analysis: disposal of graphite in containers with encapsulant and without it. It was evaluated that the maximal fractional flux of inorganic {sup 14}C into the geosphere can vary from 10{sup −} {sup 11} y{sup −} {sup 1} (for non-encapsulated graphite) to 10{sup −} {sup 12} y{sup −} {sup 1} (for encapsulated graphite) while of organic {sup 14}C it was about 10{sup −} {sup 3} y{sup −} {sup 1} of its inventory. Such difference demonstrates that investigations on the {sup 14}C inventory and chemical form in which it is released are especially important. The parameter with the highest influence on the maximal flux into the geosphere for inorganic {sup 14}C transfer was the sorption coefficient in the backfill and for organic {sup 14}C transfer – the backfill hydraulic conductivity. - Highlights: • Graphite moderated nuclear reactors are being decommissioned. • We studied interaction of disposed material with surrounding environment. • Specifically {sup 14}C transfer through engineered barriers of a geological repository. • Organic {sup 14}C flux to geosphere is considerably higher than inorganic

  19. Development of a Scanning Microscale Fast Neutron Irradiation Platform for Examining the Correlation Between Local Neutron Damage and Graphite Microstructure

    Energy Technology Data Exchange (ETDEWEB)

    Pinhero, Patrick [Univ. of Missouri, Columbia, MO (United States); Windes, William [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-10

    The fast particle radiation damage effect of graphite, a main material in current and future nuclear reactors, has significant influence on the utilization of this material in fission and fusion plants. Atoms on graphite crystals can be easily replaced or dislocated by fast protons and result in interstitials and vacancies. The currently accepted model indicates that after most of the interstitials recombine with vacancies, surviving interstitials form clusters and furthermore gather to create loops with each other between layers. Meanwhile, surviving vacancies and interstitials form dislocation loops on the layers. The growth of these inserted layers cause the dimensional increase, i.e. swelling, of graphite. Interstitial and vacancy dislocation loops have been reported and they can easily been observed by electron microscope. However, observation of the intermediate atom clusters becomes is paramount in helping prove this model. We utilize fast protons generated from the University of Missouri Research Reactor (MURR) cyclotron to irradiate highly- oriented pyrolytic graphite (HOPG) as target for this research. Post-irradiation examination (PIE) of dosed targets with high-resolution transmission electron microscopy (HRTEM) has permit observation and analysis of clusters and dislocation loops to support the proposed theory. Another part of the research is to validate M.I. Heggie’s Ruck and Tuck model, which introduced graphite layers may fold under fast particle irradiation. Again, we employed microscopy to image irradiated specimens to determine how the extent of Ruck and Tuck by calculating the number of folds as a function of dose. Our most significant accomplishment is the invention of a novel class of high-intensity pure beta-emitters for long-term lightweight batteries. We have filed four invention disclosure records based on the research conducted in this project. These batteries are lightweight because they consist of carbon and tritium and can be

  20. Irradiation creep and stress-enhanced swelling of Fe-16Cr-15Ni-Nb austenitic stainless steel in BN-350

    Energy Technology Data Exchange (ETDEWEB)

    Vorobjev, A.N.; Porollo, S.I.; Konobeev, Yu.V. [Institute of Physics and Power Engineering, Obninsk (Russian Federation)] [and others

    1997-04-01

    Irradiation creep and void swelling will be important damage processes for stainless steels when subjected to fusion neutron irradiation at elevated temperatures. The absence of an irradiation device with fusion-relevant neutron spectra requires that data on these processes be collected in surrogate devices such as fast reactors. This paper presents the response of an annealed austenitic steel when exposed to 60 dpa at 480{degrees}C and to 20 dpa at 520{degrees}C. This material was irradiated as thin-walled argon-pressurized tubes in the BN-350 reactor located in Kazakhstan. These tubes were irradiated at hoop stresses ranging from 0 to 200 MPa. After irradiation both destructive and non-destructive examination was conducted.

  1. Critical current densities and flux creep rate in Co-doped BaFe2As2 with columnar defects introduced by heavy-Ion irradiation

    International Nuclear Information System (INIS)

    Nakajima, Y.; Tsuchiya, Y.; Taen, T.; Yagyuda, H.; Tamegai, T.; Okayasu, S.; Sasase, M.; Kitamura, H.; Murakami, T.

    2010-01-01

    We report the formation of columnar defects in Co-doped BaFe 2 As 2 single crystals with different heavy-ion irradiations. The formation of columnar defects by 200 MeV Au ion irradiation is confirmed by transmission electron microscopy and their density is about 40% of the irradiation dose. Magneto-optical imaging and bulk magnetization measurements reveal that the critical current density J c is enhanced in the 200 MeV Au and 800 MeV Xe ion irradiated samples while J c is unchanged in the 200 MeV Ni ion irradiated sample. We also find that vortex creep rates are strongly suppressed by the columnar defects. We compare the effect of heavy-ion irradiation into Co-doped BaFe 2 As 2 and cuprate superconductors.

  2. Contributions for the international conference on carbon and graphite CARBON '88

    International Nuclear Information System (INIS)

    Delle, W.

    1988-08-01

    This report is the compilation of three papers prepared by the Kernforschungsanlage Juelich GmbH (KFA) in collaboration with other partners for the International Conference CARBON '88. The topics were as follows: 1.) Fracture toughness of fast neutron irradiated graphite (W. Delle, H. Derz, G. Kleist, H. Nickel, W. Thiele); 2.) The irradiation creep characteristics of graphite to high fluences (C.R. Kennedy, M. Cundy, G. Kleist); and 3.) New silicon carbide materials starting with the Coat-Mix procedure (H.K. Luhleich, K. Bach, F.J. Dias, M. Kampel, F. Koch, H. Nickel). (orig./MM)

  3. Controlling magnetic domain wall motion in the creep regime in He+-irradiated CoFeB/MgO films with perpendicular anisotropy

    International Nuclear Information System (INIS)

    Herrera Diez, L.; García-Sánchez, F.; Adam, J.-P.; Devolder, T.; Eimer, S.; El Hadri, M. S.; Ravelosona, D.; Lamperti, A.; Mantovan, R.; Ocker, B.

    2015-01-01

    This study presents the effective tuning of perpendicular magnetic anisotropy in CoFeB/MgO thin films by He + ion irradiation and its effect on domain wall motion in a low field regime. Magnetic anisotropy and saturation magnetisation are found to decrease as a function of the irradiation dose which can be related to the observed irradiation-induced changes in stoichiometry at the CoFeB/MgO interface. These changes in the magnetic intrinsic properties of the film are reflected in the domain wall dynamics at low magnetic fields (H) where irradiation is found to induce a significant decrease in domain wall velocity (v). For all irradiation doses, domain wall velocities at low fields are well described by a creep law, where Ln(v) vs. H −1∕4 behaves linearly, up to a maximum field H*, which has been considered as an approximation to the value of the depinning field H dep . In turn, H* ≈ H dep is seen to increase as a function of the irradiation dose, indicating an irradiation-induced extension of the creep regime of domain wall motion

  4. Controlling magnetic domain wall motion in the creep regime in He{sup +}-irradiated CoFeB/MgO films with perpendicular anisotropy

    Energy Technology Data Exchange (ETDEWEB)

    Herrera Diez, L., E-mail: liza.herrera-diez@ief.u-psud.fr; García-Sánchez, F.; Adam, J.-P.; Devolder, T.; Eimer, S.; El Hadri, M. S.; Ravelosona, D. [Institut d' Electronique Fondamentale, Université Paris-Sud, UMR CNRS 8622, 91405 Orsay (France); Lamperti, A.; Mantovan, R. [Laboratorio MDM, IMM-CNR, Via C. Olivetti 2, 20864 Agrate (MB) (Italy); Ocker, B. [Singulus Technology AG, Hanauer Landstrasse 103, 63796 Kahl am Main (Germany)

    2015-07-20

    This study presents the effective tuning of perpendicular magnetic anisotropy in CoFeB/MgO thin films by He{sup +} ion irradiation and its effect on domain wall motion in a low field regime. Magnetic anisotropy and saturation magnetisation are found to decrease as a function of the irradiation dose which can be related to the observed irradiation-induced changes in stoichiometry at the CoFeB/MgO interface. These changes in the magnetic intrinsic properties of the film are reflected in the domain wall dynamics at low magnetic fields (H) where irradiation is found to induce a significant decrease in domain wall velocity (v). For all irradiation doses, domain wall velocities at low fields are well described by a creep law, where Ln(v) vs. H{sup −1∕4} behaves linearly, up to a maximum field H*, which has been considered as an approximation to the value of the depinning field H{sub dep}. In turn, H* ≈ H{sub dep} is seen to increase as a function of the irradiation dose, indicating an irradiation-induced extension of the creep regime of domain wall motion.

  5. AGC-2 Specimen Post Irradiation Data Package Report

    Energy Technology Data Exchange (ETDEWEB)

    Windes, William Enoch [Idaho National Lab. (INL), Idaho Falls, ID (United States); Swank, W. David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rohrbaugh, David T. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Cottle, David L. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-08-01

    This report documents results of the post-irradiation examination material property testing of the creep, control, and piggyback specimens from the irradiation creep capsule Advanced Graphite Creep (AGC)-2 are reported. This is the second of a series of six irradiation test trains planned as part of the AGC experiment to fully characterize the neutron irradiation effects and radiation creep behavior of current nuclear graphite grades. The AGC-2 capsule was irradiated in the Idaho National Laboratory Advanced Test Reactor at a nominal temperature of 600°C and to a peak dose of 5 dpa (displacements per atom). One-half of the creep specimens were subjected to mechanical stresses (an applied stress of either 13.8, 17.2, or 20.7 MPa) to induce irradiation creep. All post-irradiation testing and measurement results are reported with the exception of the irradiation mechanical strength testing, which is the last destructive testing stage of the irradiation testing program. Material property tests were conducted on specimens from 15 nuclear graphite grades using a similar loading configuration as the first AGC capsule (AGC-1) to provide easy comparison between the two capsules. However, AGC-2 contained an increased number of specimens (i.e., 487 total specimens irradiated) and replaced specimens of the minor grade 2020 with the newer grade 2114. The data reported include specimen dimensions for both stressed and unstressed specimens to establish the irradiation creep rates, mass and volume data necessary to derive density, elastic constants (Young’s modulus, shear modulus, and Poisson’s ratio) from ultrasonic time-of-flight velocity measurements, Young’s modulus from the fundamental frequency of vibration, electrical resistivity, and thermal diffusivity and thermal expansion data from 100–500°C. No data outliers were determined after all measurements were completed. A brief statistical analysis was performed on the irradiated data and a limited comparison between

  6. Non Destructive Thermal Analysis and In Situ Investigation of Creep Mechanism of Graphite and Ceramic Composites using Phase-sensitive THz Imaging & Nonlinear Resonant Ultrasonic Spectroscopy

    International Nuclear Information System (INIS)

    Zhang, XI-Cheng; Redo-Scanchez, Albert

    2012-01-01

    In this project, we conducted a comprehensive study on nuclear graphite properties with terahertz (THz) imaging. Graphite samples from Idaho National Laboratory were carefully imaged by continuous wave (CW) THz. The CW THz imaging of graphite shows that the samples from different billet with different fabricating conditions have different pore size and structure. Based on this result, we then used a phase sensitive THz system to study the graphite properties. In this exploration, various graphite were studied. By imaging nuclear graphite samples in reflection mode at nine different incident polarization angles using THz time-domain spectroscopy, we find that different domain distributions and levels of porosity will introduce polarization dependence in THz reflectivity. Sample with higher density is less porous and has a smaller average domain distribution. As a consequence, it is less polarization-dependent and the polarization-dependent frequency is higher. The results also show that samples oxidized at higher temperatures tend to be more polarization dependent. The graphite from the external billet is more polarization dependent compared to that from the center billet. In addition, we performed laser-based ultrasonic measurements on these graphite samples. The denser, unoxidized samples allow surface acoustic waves to propagate more rapidly than in the samples that had already undergone oxidation. Therefore, for the oxidized samples, the denser samples show less polarization-dependence, higher polarization-dependent frequency, and allow the surface acoustic waves propagate faster.

  7. Direct synthesis of sp-bonded carbon chains on graphite surface by femtosecond laser irradiation

    International Nuclear Information System (INIS)

    Hu, A.; Rybachuk, M.; Lu, Q.-B.; Duley, W. W.

    2007-01-01

    Microscopic phase transformation from graphite to sp-bonded carbon chains (carbyne) and nanodiamond has been induced by femtosecond laser pulses on graphite surface. UV/surface enhanced Raman scattering spectra and x-ray photoelectron spectra displayed the local synthesis of carbyne in the melt zone while nanocrystalline diamond and trans-polyacetylene chains form in the edge area of gentle ablation. These results evidence possible direct 'writing' of variable chemical bonded carbons by femtosecond laser pulses for carbon-based applications

  8. Prediction of long-term creep curves

    International Nuclear Information System (INIS)

    Oikawa, Hiroshi; Maruyama, Kouichi

    1992-01-01

    This paper aims at discussing how to predict long-term irradiation enhanced creep properties from short-term tests. The predictive method based on the θ concept was examined by using creep data of ferritic steels. The method was successful in predicting creep curves including the tertiary creep stage as well as rupture lifetimes. Some material constants involved in the method are insensitive to the irradiation environment, and their values obtained in thermal creep are applicable to irradiation enhanced creep. The creep mechanisms of most engineering materials definitely change at the athermal yield stress in the non-creep regime. One should be aware that short-term tests must be carried out at stresses lower than the athermal yield stress in order to predict the creep behavior of structural components correctly. (orig.)

  9. IAEA International Database on Irradiated Nuclear Graphite Properties. Summary report of consultants' meeting. 12. meeting of the Technical Steering Committee

    International Nuclear Information System (INIS)

    Chung, H.K.; Wickham, A.J.

    2010-02-01

    The 12th Meeting of the Technical Steering Committee for the International Database on Irradiated Nuclear Graphite Properties was held on 12-13 November 2009 at the IAEA Headquarters, Vienna, Austria. All discussions, recommendations and actions of this Consultants' Meeting are recorded in this report. The purposes of the meeting were to review the matters and actions identified in the previous meeting, undertake a review of the current status of the database, and make recommendations for action over the next year. This report contains the status of the identified actions as well as a summary of the recommendations on enhancements to the database. (author)

  10. Creep-fatigue deformation behaviour of OFHC-copper and CuCrZr alloy with different heat treatments and with and without neutron irradiation

    International Nuclear Information System (INIS)

    Singh, B.N.; Johansen, B.S.; Li, M.; Stubbins, J.F.

    2005-08-01

    The creep-fatigue interaction behaviour of a precipitation hardened CuCrZr alloy was investigated at 295 and 573 K. To determine the effect of irradiation a number of fatigue specimens were irradiated at 333 and 573 K to a dose level in the range of 0.2 - 0.3 dpa and were tested at room temperature and 573 K, respectively. The creep-fatigue deformation behaviour of OFHC-copper was also investigated but only in the unirradiated condition and at room temperature. The creep-fatigue interaction was simulated by applying a certain holdtime on both tension and compression sides of the cyclic loading with a frequency of 0.5 Hz. Holdtimes of up to 1000 seconds were used. Creep-fatigue experiments were carried out using strain, load and extension controlled modes of cyclic loading. In addition, a number of 'interrupted' creep-fatigue tests were performed on the prime aged CuCuZr specimens in the strain controlled mode with a strain amplitude of 0.5% and a holdtime of 10 seconds. The lifetimes in terms of the number of cycles to failure were determined at different strain and load amplitudes at each holdtime. Post-deformation microstructures was investigated using a transmission electron microscopy. The main results of these investigations are presented and their implications are briefly discussed in the present report. The central conclusion emerging from the present work is that the application of holdtime generally reduces the number of cycles to failure. The largest reduction was found to be in the case of OFHC-copper. Surprisingly, the magnitude of this reduction is found to be larger at lower levels of strain or stress amplitudes, particularly when the level of the stress amplitude is below the monotonic yield strength of the material. The reduction in the yield strength due to overaging heat treatments causes a substantial decrease in the number of cycles to failure at all holdtimes investigated. The increase in the yield strength due to neutron irradiation at 333 K

  11. Innovative approaches to the Management of Irradiated Nuclear Graphite Wastes: Addressing the Challenges through International Collaboration with Project 'GRAPA'

    International Nuclear Information System (INIS)

    Wickham, A.J.; Ojovan, M.; O'Sullivan, P.; )

    2017-01-01

    There exists more than 250.000 tonnes of irradiated (and therefore radioactive) nuclear graphite (i-graphite) in the world, primarily as a result of the development of graphite-moderated power-reactor systems, initially for defence and subsequently for commercial purposes. Only a very small number of such plants have been dismantled and, for most cases, the final destiny of the irradiated graphite remains unresolved. Future high-temperature reactor programmes, such as the Chinese HTR-PM development, will produce more graphite and carbonaceous wastes from both structural components and the fuel pebbles (which are approximately 96% carbonaceous), the latter producing a continuous stream of so-called 'operational waste'. The problem of dismantling irradiated graphite reactor stacks, possibly distorted through neutron damage and in some cases degraded further by radiation-chemical attack by gaseous coolants, and then finding the appropriate treatments and final destiny of the material, has exercised both the European Union and the International Atomic Energy Agency for more than 25 years, seeking to address the different issues and available disposal solutions in different IAEA Member States. An IAEA collaborative research programme on treatment options has recently been completed, and an active group of international specialists in this area has now been established as part of the IAEA International Decommissioning Network under the envelope of Project 'GRAPA' (Irradiated Graphite Processing Approaches), which includes representatives from Belgium, China, France, Germany, India, Italy, Lithuania, Rep. of Korea, Romania, Spain, Switzerland, Ukraine and the Russian Federation with direct responsibilities for various parts of the decommissioning and graphite-disposal process in a variety of reactor designs. Interest has also been expressed by colleagues from Sweden and Japan. Work is in progress on a number of topic areas where weaknesses in the

  12. Analysis of Current HT9 Creep Correlations and Modification

    International Nuclear Information System (INIS)

    Lee, Cheol Min; Sohn, Dongseong; Cheon, Jin Sik

    2014-01-01

    It has high thermal conductivity, high mechanical strength and low irradiation induced swelling. However high temperature creep of HT9 has always been a life limiting factor. Above 600 .deg. C, the dislocation density in HT9 is decreased and the M 23 C 6 precipitates coarsen, these processes are accelerated if there is irradiation. Finally microstructural changes at high temperature lead to lower creep strength and large creep strain. For HT9 to be used as a future cladding, creep behavior of the HT9 should be predicted accurately based on the physical understanding of the creep phenomenon. Most of the creep correlations are composed of irradiation creep and thermal creep terms. However, it is certain that in-pile thermal creep and out-of-pile thermal creep are different because of the microstructure changes induced from neutron irradiation. To explain creep behavior more accurately, thermal creep contributions other than neutron irradiation should be discriminated in a creep correlation. To perform this work, existing HT9 creep correlations are analyzed, and the results are used to develop more accurate thermal creep correlation. Then, the differences between in-pile thermal creep and out-of-pile thermal creep are examined

  13. Precision rectifier detectors for ac resistance bridge measurements with application to temperature control systems for irradiation creep experiments

    Energy Technology Data Exchange (ETDEWEB)

    Duncan, M. G.

    1977-05-01

    The suitability of several temperature measurement schemes for an irradiation creep experiment is examined. It is found that the specimen resistance can be used to measure and control the sample temperature if compensated for resistance drift due to radiation and annealing effects. A modified Kelvin bridge is presented that allows compensation for resistance drift by periodically checking the sample resistance at a controlled ambient temperature. A new phase-insensitive method for detecting the bridge error signals is presented. The phase-insensitive detector is formed by averaging the magnitude of two bridge voltages. Although this method is substantially less sensitive to stray reactances in the bridge than conventional phase-sensitive detectors, it is sensitive to gain stability and linearity of the rectifier circuits. Accuracy limitations of rectifier circuits are examined both theoretically and experimentally in great detail. Both hand analyses and computer simulations of rectifier errors are presented. Finally, the design of a temperature control system based on sample resistance measurement is presented. The prototype is shown to control a 316 stainless steel sample to within a 0.15/sup 0/C short term (10 sec) and a 0.03/sup 0/C long term (10 min) standard deviation at temperatures between 150 and 700/sup 0/C. The phase-insensitive detector typically contributes less than 10 ppM peak resistance measurement error (0.04/sup 0/C at 700/sup 0/C for 316 stainless steel or 0.005/sup 0/C at 150/sup 0/C for zirconium).

  14. Brazing graphite to graphite

    International Nuclear Information System (INIS)

    Peterson, G.R.

    1976-01-01

    Graphite is joined to graphite by employing both fine molybdenum powder as the brazing material and an annealing step that together produce a virtually metal-free joint exhibiting properties similar to those found in the parent graphite. Molybdenum powder is placed between the faying surfaces of two graphite parts and melted to form molybdenum carbide. The joint area is thereafter subjected to an annealing operation which diffuses the carbide away from the joint and into the graphite parts. Graphite dissolved by the dispersed molybdenum carbide precipitates into the joint area, replacing the molybdenum carbide to provide a joint of graphite

  15. Buckle, ruck and tuck: A proposed new model for the response of graphite to neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Heggie, M.I., E-mail: m.i.heggie@sussex.ac.uk [Chemistry Subject Group, School of Life Sciences, University of Sussex, Falmer, Brighton BN1 9QJ (United Kingdom); Suarez-Martinez, I. [Nanochemistry Research Institute, Department of Chemistry, Curtin University of Technology, GPO Box U1987, Perth 6845, Western Australia (Australia); Davidson, C.; Haffenden, G. [Chemistry Subject Group, School of Life Sciences, University of Sussex, Falmer, Brighton BN1 9QJ (United Kingdom)

    2011-06-30

    The default theory of radiation damage in graphite invokes Frenkel pair formation as the principal cause of physical property changes. We set out its inadequacies and present two new mechanisms that contribute to a better account for changes in dimension and stored energy. Damage depends on the substrate temperature, undergoing a change at approximately 250 deg. C. Below this temperature particle radiation imparts a permanent, nano-buckling to the layers. Above it, layers fold, forming what we describe as a ruck and tuck defect. We present first principles and molecular mechanics calculations of energies and structures to support these claims. Necessarily we extend the dislocation theory of layered materials. We cite good experimental evidence for these features from the literature on radiation damage in graphite.

  16. Studies of mechanical properties and irradiation damage nucleation of HTGR graphites. Final report

    International Nuclear Information System (INIS)

    Thrower, P.A.

    1981-05-01

    Since the submission of the last report (COO-2712-6) work has concentrated on the examination of the effects of oxidation on the compressive strengths of graphites doped with iron, vanadium and calcium. The purpose of the investigation was to determine the relative effects of the impurities on the rates of oxidation in air, CO 2 and H 2 O and the resultant reduction in compressive strength

  17. Direct synthesis of graphitic mesoporous carbon from green phenolic resins exposed to subsequent UV and IR laser irradiations

    Science.gov (United States)

    Sopronyi, Mihai; Sima, Felix; Vaulot, Cyril; Delmotte, Luc; Bahouka, Armel; Matei Ghimbeu, Camelia

    2016-01-01

    The design of mesoporous carbon materials with controlled textural and structural features by rapid, cost-effective and eco-friendly means is highly demanded for many fields of applications. We report herein on the fast and tailored synthesis of mesoporous carbon by UV and IR laser assisted irradiations of a solution consisting of green phenolic resins and surfactant agent. By tailoring the UV laser parameters such as energy, pulse repetition rate or exposure time carbon materials with different pore size, architecture and wall thickness were obtained. By increasing irradiation dose, the mesopore size diminishes in the favor of wall thickness while the morphology shifts from worm-like to an ordered hexagonal one. This was related to the intensification of phenolic resin cross-linking which induces the reduction of H-bonding with the template as highlighted by 13C and 1H NMR. In addition, mesoporous carbon with graphitic structure was obtained by IR laser irradiation at room temperature and in very short time periods compared to the classical long thermal treatment at very high temperatures. Therefore, the carbon texture and structure can be tuned only by playing with laser parameters, without extra chemicals, as usually required. PMID:28000781

  18. Capsule development and utilization for material irradiation tests; study on the in-pile creep measuring method of zirconium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Yong; Lee, Byung Kee; Lee, Jong Jea; Kim, Chang Sik; Kim, B. Hun; Cho, I. Sik [Sunmoon University, Asan (Korea)

    2002-02-01

    The final objective of this project is to obtain a design and fabrication technology of an in-pile creep test machine of zirconium alloys. First, design concepts of the in-pile creep test machines of various foreign countries were reviewed and a preliminary design of the equipment was carried. Second, the mock-up of the in-pile creep test machine was fabricated based on the preliminary design. The mock-up consisted of upper and lower grips, a yoke, a pressure chamber including a bellows, a push rod and LVDT. Each part was made of 304 L stainless steel. The average surface roughness of the parts was 1.0-14.7 {mu}m. The mock-up precisely determined an extension of a specimen by gas pressure. Finally, in-pile creep capsule was designed, fabricated and modified. High pure aluminum blocks were put in the capsule. Considering heat transfer coefficients of helium and nitrogen gases, the cooling efficiency is about 4 .deg. C at the condition of 300 .deg. C creep test. Yield strength, ultimate tensile strength and elongation at 300 .deg. C were 335 MPa, 591 MPa, 19.8%, respectively. which were lower than the values at room temperature, 353 MPa, 740 MPa, 12.5%. This study gave an important technology related to design, fabrication and performance tests of the in-pile creep test machine, which is applied to the fabrication of a special capsule and also used for the fundamental data for the fabrication of various in-pile creep capsules. 6 refs., 45 figs., 5 tabs. (Author)

  19. Analysis of Wigner energy release process in graphite stack of shut-down uranium-graphite reactor

    OpenAIRE

    Bespala, E. V.; Pavliuk, A. O.; Kotlyarevskiy, S. G.

    2015-01-01

    Data, which finding during thermal differential analysis of sampled irradiated graphite are presented. Results of computational modeling of Winger energy release process from irradiated graphite staking are demonstrated. It's shown, that spontaneous combustion of graphite possible only in adiabatic case.

  20. Production of an impermeable composite of irradiated graphite and glass by hot isostatic pressing as a long term leach resistant waste form

    Energy Technology Data Exchange (ETDEWEB)

    Fachinger, Johannes; Muller, Walter [FNAG ZU Hanau, Hanau (Germany); Marsat, Eric [FNAG SAS Le Pont de Claix (France); Grosse, Karl-Heinz; Seemann, Richard [ALD Hanau (Germany); Scales, Charlie; Easton, Michael Mark [NNL, Workington (United Kingdom); Anthony Banford [NNL, Warrington (United Kingdom); University of Manchester, Manchester (United Kingdom)

    2013-07-01

    Around 250,000 tons of irradiated graphite (i-graphite) exists worldwide and can be considered as a current waste or future waste stream. The largest national i-graphite inventory is located in UK (∼ 100,000 tons) with significant quantities also in Russia and France [5]. Most of the i-graphite remains in the cores of shutdown nuclear reactors including the MAGNOX type in UK and the UNGG in France. Whilst there are still operational power reactors with graphite cores, such as the Russian RBMKs and the AGRs in UK, all of them will reach their end of life during the next two decades. The most common reference waste management option of i-graphite is a wet or dry retrieval of the graphite blocks from the reactor core and the grouting of these blocks in a container without further conditioning. This produces large waste package volumes because the encapsulation capacity of the grout is limited and large cavities in the graphite blocks could reduce the packing densities. Packing densities from 0.5 to 1 tons per cubic meter have been assumed for grouting solutions. Furthermore the grout is permeable. This could over time allow the penetration of aqueous phases into the waste block and a potential dissolution and release of radionuclides. As a result particularly highly soluble radionuclides may not be retained by the grout. Vitrification could present an alternative, however a similar waste package volume increase may be expected since the encapsulation capacity of glass is potentially similar to or worse than that of grout. FNAG has developed a process for the production of a graphite-glass composite material called Impermeable Graphite Matrix (IGM) [3]. This process is also applicable to irradiated graphite which allows the manufacturing of an impermeable material without volume increase. Crushed i-graphite is mixed with 20 vol.% of glass and then pressed under vacuum at an elevated temperature in an axial hot vacuum press (HVP). The obtained product has zero or

  1. Production of an impermeable composite of irradiated graphite and glass by hot isostatic pressing as a long term leach resistant waste form

    International Nuclear Information System (INIS)

    Fachinger, Johannes; Muller, Walter; Marsat, Eric; Grosse, Karl-Heinz; Seemann, Richard; Scales, Charlie; Easton, Michael Mark; Anthony Banford

    2013-01-01

    Around 250,000 tons of irradiated graphite (i-graphite) exists worldwide and can be considered as a current waste or future waste stream. The largest national i-graphite inventory is located in UK (∼ 100,000 tons) with significant quantities also in Russia and France [5]. Most of the i-graphite remains in the cores of shutdown nuclear reactors including the MAGNOX type in UK and the UNGG in France. Whilst there are still operational power reactors with graphite cores, such as the Russian RBMKs and the AGRs in UK, all of them will reach their end of life during the next two decades. The most common reference waste management option of i-graphite is a wet or dry retrieval of the graphite blocks from the reactor core and the grouting of these blocks in a container without further conditioning. This produces large waste package volumes because the encapsulation capacity of the grout is limited and large cavities in the graphite blocks could reduce the packing densities. Packing densities from 0.5 to 1 tons per cubic meter have been assumed for grouting solutions. Furthermore the grout is permeable. This could over time allow the penetration of aqueous phases into the waste block and a potential dissolution and release of radionuclides. As a result particularly highly soluble radionuclides may not be retained by the grout. Vitrification could present an alternative, however a similar waste package volume increase may be expected since the encapsulation capacity of glass is potentially similar to or worse than that of grout. FNAG has developed a process for the production of a graphite-glass composite material called Impermeable Graphite Matrix (IGM) [3]. This process is also applicable to irradiated graphite which allows the manufacturing of an impermeable material without volume increase. Crushed i-graphite is mixed with 20 vol.% of glass and then pressed under vacuum at an elevated temperature in an axial hot vacuum press (HVP). The obtained product has zero or

  2. Irradiation creep and swelling of AISI 316 to exposures of 130 dpa at 385?400$deg;C

    Science.gov (United States)

    Garner, F. A.; Porter, D. L.

    1988-07-01

    The creep and swelling of AISI 316 stainless steel have been studied at 385 to 400°C in EBR-II to doses of 130 dpa. Most creep capsules were operated at constant stress and temperature but mid-life changes in these variables were also made. This paper concentrates on the behavior of the 20% cold-worked condition but five other conditions were also studied. Swelling at ⩽ 400° C was found to lose the sensitivity to stress exhibited at higher temperatures while the creep rate was found to retain linear dependencies on both stress and swelling rate. The creep coefficients extracted at 400°C agree with those found in other experiments conducted at higher temperatures. In the temperature range of ⩽ 400° C, swelling is in the recombinationdominated regime and the swelling rate falls strongly away from the ~1%/dpa rate observed at higher temperatures. These lower rates of creep and swelling, coupled with the attainment of high damage levels without failure, encourage the use of AISI 316 in the construction of water-cooled fusion first walls operating at temperatures below 400°C.

  3. The influence of Boron on creep-rupture behaviour of austenitic unstabilized and Nb-stabilized stainless steel X8CrNi 1613 in unirradiated and irradiated condition

    International Nuclear Information System (INIS)

    Sen, Susant Kumar.

    1976-10-01

    The present study deals with influence of boron on creep-rupture behaviour in unirradiated condition at 650 0 C along with precipitation behaviour, heat-treatment and recrystallization of unstabilized and stabilized steel. The results of creep-rupture tests on unirradiated specimens show that boron exerts a beneficial effect on the rupture life and ductility. Boron losses its beneficial effect on creep properties in unstabilized steel by prolong creeping. The magnitude of beneficial effect of Boron on creep properties depends upon the initial boron distribution which influences the number, size and distribution of the precipitates. Boron promotes the precipitation of type M 23 C 6 Carbides in the grain as well as at the grain boundary. Boron segregates in atomic form during slow cooling from austenitizing temperature. The recrystallization will be delayed by the presence of boron. The results of creep tests at 650 0 C shows that boron exerts a beneficial effect on creep life of irradiated steels. (orig./GSC) [de

  4. Toxicological characterization of chemicals produced from laser irradiation of graphite composite materials

    International Nuclear Information System (INIS)

    Kwan, J.

    1990-11-01

    One of the major potential hazards associated with laser machining of graphite composite materials is the toxic fumes and gases that are generated. When exposed to the intense energy of the laser beam, the organic polymer matrix of the composite material may decompose into various toxic by-products. To advance the understanding of the laser machining process from a health and safety viewpoint, this particular study consisted of the following steps: collect and analyze gaseous by-products generated during laser machining; collect particulates generated during laser machining and chemically extract them to determine the chemical species that may have absorbed or recondensed onto these particles; and review and evaluate the toxicity of the identified chemical species

  5. Nuclear graphite ageing and turnaround

    International Nuclear Information System (INIS)

    Marsden, B.J.; Hall, G.N.; Smart, J.

    2001-01-01

    Graphite moderated reactors are being operated in many countries including, the UK, Russia, Lithuania, Ukraine and Japan. Many of these reactors will operate well into the next century. New designs of High Temperature Graphite Moderated Reactors (HTRS) are being built in China and Japan. The design life of these graphite-moderated reactors is governed by the ageing of the graphite core due to fast neutron damage, and also, in the case of carbon dioxide cooled reactors by the rate of oxidation of the graphite. Nuclear graphites are polycrystalline in nature and it is the irradiation-induced damage to the individual graphite crystals that determines the material property changes with age. The life of a graphite component in a nuclear reactor can be related to the graphite irradiation induced dimensional changes. Graphites typically shrink with age, until a point is reached where the shrinkage stops and the graphite starts to swell. This change from shrinkage to swelling is known as ''turnaround''. It is well known that pre-oxidising graphite specimens caused ''turnaround'' to be delayed, thus extending the life of the graphite, and hence the life of the reactor. However, there was no satisfactory explanation of this behaviour. This paper presents a numerical crystal based model of dimensional change in graphite, which explains the delay in ''turnaround'' in the pre-oxidised specimens irradiated in a fast neutron flux, in terms of crystal accommodation and orientation and change in compliance due to radiolytic oxidation. (author)

  6. Graphite targets at LAMPF

    International Nuclear Information System (INIS)

    Brown, R.D.; Grisham, D.L.

    1983-01-01

    Rotating polycrystalline and stationary pyrolytic graphite target designs for the LAMPF experimental area are described. Examples of finite element calculations of temperatures and stresses are presented. Some results of a metallographic investigation of irradiated pyrolytic graphite target plates are included, together with a brief description of high temperature bearings for the rotating targets

  7. Interpretation of measurements made by oscillations of irradiated fuels in natural uranium, graphite-gas piles

    International Nuclear Information System (INIS)

    Laponche, Bernard; Luffin, Jean; Brunet, Max; Guerange, Jacques; Tonolli, Jacky

    1969-06-01

    When considering a pile operation, it is interesting to know the evolution of fuel quality with respect to irradiation, i.e. the variation of its fission rate and of its absorption rate. In order to experimentally obtain these features, a method is to introduce an irradiated cartridge into a critical reactor and to measure the induced effect on its reactivity and on the neutron density at the vicinity of the cartridge. An oscillation method presented in another document and based on a periodic introduction of fuel sample into a critical reactor allows, from the measurement of reactivity variation (global signal), and of the neutron density (local signal), effective macroscopic fission and absorption cross sections of this sample to be obtained. As previous studies revealed that the interpretation of the local signal was notably delicate, this information has been replaced by computed information, the fission rate, which is determined by means of the COREGRAF1 code. Thus, the remaining quantity to be obtained is the fuel absorption rate. The authors report studies performed on several sets of cartridges from different reactors, and with an irradiation range from about 700 to 4000 MWJ/T. In a first part, they describe the characteristics of the studied cartridges, their irradiation and measurement conditions, and the use of the evolution code. In a second part, they try to define the interpretation of oscillation-based measurements by using two methods, a first and fast one which gives an approximation of results, and a more elaborated second one which complies with measurement conditions. The last part presents and discusses the obtained results [fr

  8. PRELIMINARY RESULTS OF THE AGC-4 IRRADIATION IN THE ADVANCED TEST REACTOR AND DESIGN OF AGC-5 (HTR16-18469)

    Energy Technology Data Exchange (ETDEWEB)

    Davenport, Michael; Petti, D. A.

    2016-11-01

    The United States Department of Energy’s Advanced Reactor Technologies (ART) Program will irradiate up to six nuclear graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The graphite experiments are being irradiated over an approximate eight year period to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Very High Temperature Gas Reactor (VHTR), as well as other future gas reactors. The experiments each consist of a single capsule that contain six stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens are not be subjected to a compressive load during irradiation. The six stacks have differing compressive loads applied to the top half of diametrically opposite pairs of specimen stacks. A seventh specimen stack in the center of the capsule does not have a compressive load. The specimens are being irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There are also samples taken of the sweep gas effluent to measure any oxidation or off-gassing of the specimens that may occur during initial start-up of the experiment. The first experiment, AGC-1, started its irradiation in September 2009, and the irradiation was completed in January 2011. The second experiment, AGC-2, started its irradiation in April 2011 and completed its irradiation in May 2012. The third experiment, AGC-3, started its irradiation in late November 2012 and completed in the April of 2014. AGC-4 is currently being irradiated in the ATR. This paper will briefly discuss the preliminary irradiation results

  9. Creep rupture behavior of unidirectional advanced composites

    Science.gov (United States)

    Yeow, Y. T.

    1980-01-01

    A 'material modeling' methodology for predicting the creep rupture behavior of unidirectional advanced composites is proposed. In this approach the parameters (obtained from short-term tests) required to make the predictions are the three principal creep compliance master curves and their corresponding quasi-static strengths tested at room temperature (22 C). Using these parameters in conjunction with a failure criterion, creep rupture envelopes can be generated for any combination of in-plane loading conditions and ambient temperature. The analysis was validated experimentally for one composite system, the T300/934 graphite-epoxy system. This was done by performing short-term creep tests (to generate the principal creep compliance master curves with the time-temperature superposition principle) and relatively long-term creep rupture tensile tests of off-axis specimens at 180 C. Good to reasonable agreement between experimental and analytical results is observed.

  10. A 3-D inelastic analysis of HTR graphite structures and a comparison with A 2-D approach

    International Nuclear Information System (INIS)

    Willaschek, J.

    1979-01-01

    In High Temperature Reactor Cores (HTR) a large number of elements are constructed of nuclear graphite. The dimensions of the graphite components are limited by stresses and strains resulting from thermal loads, irradiation induced dimensional changes and stress-dependent irradiation creep. Therefore it is necessary to examine the feasibility of design concepts with regard to the structural integrity of the material. This paper presents an analysis of a radial reflector concept for use in a 3000 MWth HTR for process heat production. This concept of a pebble bed reactor (OTTO cycle) requires reflector dimensions and shapes which have previously not been used and which may exceed acceptable stress limits. Graphite reflector elements in a HTR are subject to a high fluence of fast neutrons. The fluence varies spatially within an element. Irradiation-induced strains occur which in turn vary non-linearly with the fluence. At low fluences the graphite shrinks. With increasing fluence shrinkage is saturated and after a 'turn-around' point the graphite begins to swell. The net effect of fluence gradient and irradiation-induced strain is a 'necking' of the element which moves radially outwards with time. In this paper a three-dimensional inelastic analysis of a graphite block with the above deformation history is described. The influence of irradiation on dimensional stability and other material properties was taken into account. Numerical results were obtained with the finite-element computer code ADINA, modified at INTERATOM for the task in hand. The radial reflector block was modelled using 21-node three-dimensional continuum elements of elastic-creep material. The element stiffness matrices were calculated using the standard 2x2x2 Gauss integration; material nonlinearities with quadratic displacement functions and linearised initial strains were employed. (orig.)

  11. Comparison of 3 MeV C+ Ion-Irradiation Effects between The Nuclear Graphites made of Pitch and Petroleum Cokes

    International Nuclear Information System (INIS)

    Se-Hwan, Chi; Gen-Chan, Kim; Jong-Hwa, Chang

    2006-01-01

    Currently, all the commercially available nuclear graphite grades are being made from two different cokes, i.e., petroleum coke or coal-tar pitch coke, and a coal-tar pitch binder. Of these, since the coke composes most of the graphite volume, i.e., > 70 %, it is understood that a physical, chemical, thermal, and mechanical property as well as an irradiation-induced property change will be strongly dependent on the type of coke. To obtain first-hand information on the effects of the coke type, i.e., petroleum or pitch, on the irradiation sensitivity of graphite, specimens made of IG-110 of petroleum coke and IG-430 of pitch coke were irradiated up to ∼ 19 dpa by 3 MeV C + at room temperature, and the irradiation-induced changes in the hardness, Young's modulus, Raman spectrum, and oxidation properties were characterized. Results of the TEM show that the size and density of the Mrozowski cracks appeared to be far larger and higher in the IG-110 than the IG-430. Results of the hardness test revealed a slightly higher increase in the IG-430 than the IG-110 by around 10 dpa, and the Raman spectrum measurement showed a higher (FWHM) D /(FWHM) G value for IG-430 for 0.02 ∼ 0.25 dpa. Both the hardness and Raman measurement may imply a higher irradiation sensitivity of the IG-430 than the IG-110. Results of the Young's modulus measurements showed a large data scattering, which prevented us from estimating the differences between the grades. Oxidation experiments using a TG-DTA under a flow of dry air/He = 2.5 % (flow rate: 40 CC/min) at 750 and 1000 deg C show that the IG-110 of the petroleum coke exhibits a far higher oxidation rate than the IG-430. The discrepancy between the oxidation rate of the two grades increased with an increase in the oxidation temperature and the dose. Oxidized surface pore area was larger for IG-110. Judging from the results obtained from the present experimental conditions, the irradiation sensitivity appeared to be dependent on the degree

  12. Fabrication of SnO2-Reduced Graphite Oxide Monolayer-Ordered Porous Film Gas Sensor with Tunable Sensitivity through Ultra-Violet Light Irradiation

    Science.gov (United States)

    Xu, Shipu; Sun, Fengqiang; Yang, Shumin; Pan, Zizhao; Long, Jinfeng; Gu, Fenglong

    2015-01-01

    A new graphene-based composite structure, monolayer-ordered macroporous film composed of a layer of orderly arranged macropores, was reported. As an example, SnO2-reduced graphite oxide monolayer-ordered macroporous film was fabricated on a ceramic tube substrate under the irradiation of ultra-violet light (UV), by taking the latex microsphere two-dimensional colloid crystal as a template. Graphite oxide sheets dispersed in SnSO4 aqueous solution exhibited excellent affinity with template microspheres and were in situ incorporated into the pore walls during UV-induced growth of SnO2. The growing and the as-formed SnO2, just like other photocatalytic semiconductor, could be excited to produce electrons and holes under UV irradiation. Electrons reduced GO and holes adsorbed corresponding negative ions, which changed the properties of the composite film. This film was directly used as gas-sensor and was able to display high sensitivity in detecting ethanol gas. More interestingly, on the basis of SnO2-induced photochemical behaviours, this sensor demonstrated tunable sensitivity when UV irradiation time was controlled during the fabrication process and post in water, respectively. This study provides efficient ways of conducting the in situ fabrication of a semiconductor-reduced graphite oxide film device with uniform surface structure and controllable properties. PMID:25758292

  13. Graphitic carbon nanospheres: A Raman spectroscopic investigation of thermal conductivity and morphological evolution by pulsed laser irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Agarwal, Radhe; Sahoo, Satyaprakash, E-mail: satya504@gmail.com, E-mail: rkatiyar@hpcf.upr.edu; Chitturi, Venkateswara Rao; Katiyar, Ram S., E-mail: satya504@gmail.com, E-mail: rkatiyar@hpcf.upr.edu [Department of Physics, University of Puerto Rico, San Juan, Puerto Rico 00936-8377 (United States)

    2015-12-07

    Graphitic carbon nanospheres (GCNSs) were prepared by a unique acidic treatment of multi-walled nanotubes. Spherical morphology with a narrow size distribution was confirmed by transmission electron microscopy studies. The room temperature Raman spectra showed a clear signature of D- and G-peaks at around 1350 and 1591 cm{sup −1}, respectively. Temperature dependent Raman scattering measurements were performed to understand the phonon dynamics and first order temperature coefficients related to the D- and G-peaks. The temperature dependent Raman spectra in a range of 83–473 K were analysed, where the D-peak was observed to show a red-shift with increasing temperature. The relative intensity ratio of D- to G-peaks also showed a significant rise with increasing temperature. Such a temperature dependent behaviour can be attributed to lengthening of the C-C bond due to thermal expansion in material. The estimated value of the thermal conductivity of GCNSs ∼0.97 W m{sup −1} K{sup −1} was calculated using Raman spectroscopy. In addition, the effect of pulsed laser treatment on the GCNSs was demonstrated by analyzing the Raman spectra of post irradiated samples.

  14. Temperature and irradiation effects on the behaviour of 14C and its precursor 14N in nuclear graphite. Study of a decontamination process using steam reforming

    International Nuclear Information System (INIS)

    Silbermann, Gwennaelle

    2013-01-01

    The dismantling of UNGG reactors in France will generate about 23 000 tons of radioactive graphite wastes. To manage these wastes, the radiological inventory and data on radionuclides (RN) location and speciation should be determined. 14 C was identified as an important RN for disposal due to its high initial activity and the risk of release of a mobile organic fraction in environment, after water ingress into the disposal. Hence, the objective of this thesis, carried out in partnership with EDF is to implement experimental studies to simulate and evaluate the impact of temperature, irradiation and graphite radiolytic corrosion on the in reactor behavior of 14 C and its precursor, 14 N. The obtained data are then used to study the thermal decontamination of graphite in presence of water vapor. The experimental approach aims at simulating the presence of 14 C and 14 N by the respective ion implantation of 13 C and 14 N or 15 N in virgin graphite. This study shows that, in the temperature range reached during reactor operation, (100-500 C) and without radiolytic corrosion, 13 C is thermally stable whatever the initial graphite structure. Moreover, irradiation experiments were performed on heated graphite (500 C) put in contact with a gas representative of the radiolized coolant gas. They show the synergistic role played by the oxidative species and the graphite structure disorder on the enhancement of 13 C mobility resulting in the gasification of the graphite surface and/or the selective oxidation of 13 C more weakly bound than 12 C. Concerning the pristine nitrogen, we showed first that the surface concentration reaches several hundred ppm (≤500 ppm at) and decreases at deeper depths to about 160 ppm at.. Unlike implanted 13 C, implanted nitrogen migrates at 500 C when the graphite is highly disordered (about 8 dpa) while remaining stable for a lower disorder rate (0.14 dpa). Experiments also show the synergistic role by electronic excitations and temperature

  15. Point defects and the creep of metals

    International Nuclear Information System (INIS)

    Nichols, F.A.

    1976-01-01

    Basic concepts felt to be important in diffusion-controlled creep of metals are reviewed and it is suggested that such creep is controlled by edge-dislocation climb under a rather wide range of conditions. The effect of a damage-producing flux on such creep processes is explored. It is shown that processes such as Herring-Nabarro creep are unaffected by irradiation. Evidence is presented for a climb-plus-glide mechanism of radiation creep for stresses above unirradiated yield or flow stresses. At lower stresses a preferential dislocation loop nucleation model is suggested

  16. Graphite selection for the PBMR reflector

    International Nuclear Information System (INIS)

    Marsden, B.J.; Preston, S.D.

    2000-01-01

    A high temperature, direct cycle gas turbine, graphite moderated, helium cooled, pebble-bed reactor (PBMR) is being designed and constructed in South Africa. One of the major components in the PBMR is the graphite reflector, which must be designed to last thirty-five full power years. Fast neutron irradiation changes the dimensions and material properties of reactor graphite, thus for design purposes a suitable graphite database is required. Data on the effect of irradiation on nuclear graphites has been gathered for many years, at considerable financial cost, but unfortunately these graphites are no longer available due to rationalization of the graphite industry and loss of key graphite coke supplies. However, it is possible, using un-irradiated graphite materials properties and knowledge of the particular graphite microstructure, to determine the probable irradiation behaviour. Three types of nuclear graphites are currently being considered for the PBMR reflector: an isostatically moulded, fine grained, high strength graphite and two extruded medium grained graphites of moderately high strength. Although there is some irradiation data available for these graphites, the data does not cover the temperature and dose range required for the PBMR. The available graphites have been examined to determine their microstructure and some of the key material properties are presented. (authors)

  17. Creep testing of nodular iron at ambient and elevated temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Martinsson, Aasa; Andersson-Oestling, Henrik C.M.; Seitisleam, Facredin; Wu, Rui; Sandstroem, Rolf (Swerea KIMAB AB, Stockholm (Sweden))

    2010-12-15

    The creep strain at room temperature, 100 and 125 deg C has been investigated for the ferritic nodular cast iron insert intended for use as the load-bearing part of canisters for long term disposal of spent nuclear fuel. The microstructure consisted of ferrite, graphite nodules of different sizes, compacted graphite and pearlite. Creep tests have been performed for up to 41,000 h. The specimens were cut out from material taken from two genuine inserts, I30 and I55. After creep testing, the specimens from the 100 deg C tests were hardness tested and a metallographic examination was performed. Creep strains at all temperatures appear to be logarithmic, and accumulation of creep strain diminishes with time. The time dependence of the creep strain is consistent to the W-model for primary creep. During the loading plastic strains up to 1% appeared. The maximum recorded creep strain after the loading phase was 0.025%. This makes the creep strains technically insignificant. Acoustic emission recordings during the loading of the room temperature tests showed no sounds or other evidence of microcracking during the loading phase. There is no evidence that the hardness or the graphite microstructure changed during the creep tests

  18. Development of irradiation rig in HTTR and dosimetry method. I-I type irradiation equipment

    Energy Technology Data Exchange (ETDEWEB)

    Shibata, Taiju; Kikuchi, Takayuki [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Miyamoto, Satoshi; Ogura, Kazutomo [Japan Atomic Power Co., Tokyo (Japan)

    2002-12-01

    The High Temperature Engineering Test Reactor (HTTR) is a graphite-moderated, helium gas-cooled test reactor with a maximum power of 30 MW. The HTTR aims not only to establish and upgrade the technological basis for the HTGRs but also to perform the innovative basic research on high temperature engineering with high temperature irradiation fields. It is planned that the HTTR is used to perform various engineering tests such as the safety demonstration test, high temperature test operation and irradiation test with large irradiation fields at high temperatures. This paper describes the design of the I-I type irradiation equipment developed as the first irradiation rig for the HTTR and does the planned dosimetry method at the first irradiation test. It was developed to perform in-pile creep test on a stainless steel with large standard size specimens in the HTTR. It can give great loads on the specimens stably and can control the irradiation temperature precisely. The in-core creep properties on the specimens are measured by newly developed differential transformers and the irradiation condition in the core is monitored by thermocouples and self-powered neutron detectors (SPNDs), continuously. The irradiated neutron fluence is assessed by neutron fluence monitors of small metallic wires after the irradiation. The obtained data at the first irradiation test can strongly be contributed to upgrade the technological basis for the HTGRs, since it is the first direct measurement of the in-core irradiation environments of the HTTR. (author)

  19. Comparison of 3 MeV C+ ion-irradiation effects between the nuclear graphites made of pitch and petroleum cokes

    International Nuclear Information System (INIS)

    Chi, Se-Hwan; Kim, Gen-Chan

    2008-01-01

    Three million electron volt C + irradiation effects on the microstructure (crystallinity, crystal size), mechanical properties (hardness, Young's modulus) and oxidation of IG-110 (petroleum coke) and IG-430 (pitch coke) nuclear graphites were compared based on the materials characteristics (degree of graphitization (DOG), density, porosity, type of coke, Mrozowski cracks) of the grades and the ion-irradiation conditions. The specimens were irradiated up to ∼19 dpa at room temperature. Differences in the as-received microstructure were examined by Raman spectroscopy, X-ray diffraction (XRD), optical microscope (OM) and transmission electron microscope (TEM). The ion-induced changes in the microstructure, mechanical properties and oxidation characteristics were examined by the Raman spectroscopy, microhardness and Young's modulus measurements, and scanning electron microscope (SEM). Results of the as-received microstructure condition show that the DOG of the grades appeared the same at 0.837. The size of Mrozowski cracks appeared larger in the IG-110 of the higher open and total porosity than the IG-430. After an irradiation, the changes in the crystallinity and the crystallite size, both estimated by the Raman spectrum parameters, appeared large for the IG-430 and the IG-110, respectively. The hardness had increased after an irradiation, but, the hardness increasing behaviors were reversed at around 14 dpa. Thus, the IG-430 showed a higher increase before 14 dpa, but the IG-110 showed a higher increase after 14 dpa. No-clear differences in the increase of the Young's modulus were observed between the grades mainly due to a scattering in the measurements results. The IG-110 showed a higher oxidation rate than the IG-430 both before and after an irradiation. Besides the density and porosity, a possible contribution of the well-developed Mrozowski cracks in the IG-110 was noted for the observation. All the comparisons show that, even when the differences between the

  20. Comparison of 3 MeV C + ion-irradiation effects between the nuclear graphites made of pitch and petroleum cokes

    Science.gov (United States)

    Chi, Se-Hwan; Kim, Gen-Chan

    2008-10-01

    Three million electron volt C + irradiation effects on the microstructure (crystallinity, crystal size), mechanical properties (hardness, Young's modulus) and oxidation of IG-110 (petroleum coke) and IG-430 (pitch coke) nuclear graphites were compared based on the materials characteristics (degree of graphitization (DOG), density, porosity, type of coke, Mrozowski cracks) of the grades and the ion-irradiation conditions. The specimens were irradiated up to ˜19 dpa at room temperature. Differences in the as-received microstructure were examined by Raman spectroscopy, X-ray diffraction (XRD), optical microscope (OM) and transmission electron microscope (TEM). The ion-induced changes in the microstructure, mechanical properties and oxidation characteristics were examined by the Raman spectroscopy, microhardness and Young's modulus measurements, and scanning electron microscope (SEM). Results of the as-received microstructure condition show that the DOG of the grades appeared the same at 0.837. The size of Mrozowski cracks appeared larger in the IG-110 of the higher open and total porosity than the IG-430. After an irradiation, the changes in the crystallinity and the crystallite size, both estimated by the Raman spectrum parameters, appeared large for the IG-430 and the IG-110, respectively. The hardness had increased after an irradiation, but, the hardness increasing behaviors were reversed at around 14 dpa. Thus, the IG-430 showed a higher increase before 14 dpa, but the IG-110 showed a higher increase after 14 dpa. No-clear differences in the increase of the Young's modulus were observed between the grades mainly due to a scattering in the measurements results. The IG-110 showed a higher oxidation rate than the IG-430 both before and after an irradiation. Besides the density and porosity, a possible contribution of the well-developed Mrozowski cracks in the IG-110 was noted for the observation. All the comparisons show that, even when the differences between the

  1. Decolorizing of azo dye Reactive red 24 aqueous solution using exfoliated graphite and H2O2 under ultrasound irradiation.

    Science.gov (United States)

    Li, Mei; Li, Ji-Tai; Sun, Han-Wen

    2008-07-01

    At its natural pH (6.95), the decolorization of Reactive red 24 in ultrasound, ultrasound/H2O2, exfoliated graphite, ultrasound/exfoliated graphite, exfoliated graphite/H2O2 and ultrasound/exfoliated graphite/H2O2 systems were compared. An enhancement was observed for the decolorization in ultrasound/exfoliated graphite/H2O2 system. The effect of solution pH, H2O2 and exfoliated graphite dosages, and temperature on the decolorization of Reactive red 24 was investigated. The sonochemical treatment in combination with exfoliated graphite/H2O2 showed a synergistic effect for the decolorization of Reactive red 24. The results indicated that under proper conditions, there was a possibility to remove Reactive red 24 very efficient from aqueous solution. The decolorization of other azo dyes (Reactive red 2, Methyl orange, Acid red 1, Acid red 73, Acid red 249, Acid orange 7, Acid blue 113, Acid brown 75, Acid green 20, Acid yellow 42, Acid mordant brown 33, Acid mordant yellow 10 and Direct green 1) was also investigated, at their natural pH.

  2. Nuclear graphite waste management. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    2001-05-01

    The purpose of the seminar was to bring together the specialists dealing with various aspects of radioactive graphite waste management to exchange and review information on the decommissioning, characterisation, processing and disposal of irradiated graphite from reactor cores and other graphite waste associated with reactor operation. The seminar covered radioactive graphite characterisation, the effect of irradiation on graphite components, Wigner energy, radioactive graphite waste treatment, conditioning, interim storage and long term disposal options. Individual papers presented at the seminar were indexed separately

  3. Decoupling and tuning competing effects of different types of defects on flux creep in irradiated YBa2Cu3O7-δ coated conductors

    Energy Technology Data Exchange (ETDEWEB)

    Eley, S.; Leroux, M.; Rupich, M. W.; Miller, D. J.; Sheng, H.; Niraula, P. M.; Kayani, A.; Welp, U.; Kwok, W. -K.; Civale, L.

    2016-11-15

    YBa2Cu3O7-δ coated conductors (CCs) have achieved high critical current densities (J c) that can be further increased through the introduction of additional defects using particle irradiation. However, these gains are accompanied by increases in the flux creep rate, a manifestation of competition between the different types of defects. Here, we study this competition to better understand how to design pinning landscapes that simultaneously increase J c and reduce creep. CCs grown by metal organic deposition show non-monotonic changes in the temperature-dependent creep rate, S(T). Notably, in low fields, there is a conspicuous dip to low S as the temperature (T) increases from ~20 to ~65 K. Oxygen-, proton-, and Au-irradiation substantially increase S in this temperature range. Focusing on an oxygen-irradiated CC, we investigate the contribution of different types of irradiation-induced defects to the flux creep rate. Specifically, we study S(T) as we tune the relative density of point defects to larger defects by annealing both an as-grown and an irradiated CC in O2 at temperatures T A = 250 °C–600 °C. We observe a steady decrease in S(T > 20 K) with increasing T A, unveiling the role of pre-existing nanoparticle precipitates in creating the dip in S(T) and point defects and clusters in increasing S at intermediate temperatures.

  4. Evaluation of plasma disruption simulating short pulse laser irradiation experiments on boronated graphites and CFCs [carbon fibre composites

    International Nuclear Information System (INIS)

    Stad, R.C.L. van der; Klippel, H.T.; Kraaij, G.J.

    1992-12-01

    New experimental and numerical results from disruption heat flux simulations in the millisecond range with laser beams are discussed. For a number of graphites, boronated graphites and carbon fibre composites, the effective enthalpy of ablation is determined as 30 ± 3 MJ/kg, using laser pulses of about -.3 ms. The numerical results predict the experimental results rather well. No effect of boron doping on the ablation enthalpy is found. (author). 9 refs., 4 figs., 1 tab

  5. Summary report of consultants meeting on IAEA International Database on Irradiated Nuclear Graphite Properties. 11. meeting of the Technical Steering Committee

    International Nuclear Information System (INIS)

    Humbert, D.; Wickham, A.J.

    2009-05-01

    The 11th Meeting of the Technical Steering Committee for the International Database on Irradiated Nuclear Graphite Properties was held on 25-26 March 2009 at the IAEA Headquarters, Vienna, Austria. All discussions, recommendations and actions of this Consultants' Meeting are recorded in this report. The purposes of the meeting were to review the matters and actions identified in the previous meeting, undertake a review of the current status of the database, and make recommendations for action over the next year. This report contains the status of the identified actions as well as a summary of the recommendations on enhancements to the database. (author)

  6. Summary report of consultants' meeting - IAEA International Database on Irradiated Nuclear Graphite Properties. 8th meeting of the Technical Steering Committee

    International Nuclear Information System (INIS)

    Humbert, D.; Wickham, A.J.

    2006-05-01

    The '8th Meeting of the Technical Steering Committee for the International Database on Irradiated Nuclear Graphite Properties' was held on 15-16 March 2006 at the IAEA Headquarters, Vienna, Austria. All discussions, recommendations and actions of this Consultants' Meeting are recorded in this report. The purposes of the meeting were to review the matters and actions identified in the previous meeting, undertake a review of the current status of the database and make recommendations for actions for the next year. This report contains the current status of the identified actions as well as a summary of the recommendations on enhancements to the database. (author)

  7. Structure and properties of combined coatings on C (graphite)/Al/Al2O3 base after Ti ion implantation with subsequent electron beam irradiation

    International Nuclear Information System (INIS)

    Pogrebnjak, A.D.; Pogrebnjak, N.A.; Gritsenko, B.P.; Kylyshkanov, M.K.; Ruzimov, Sh.M.

    2004-01-01

    Full text: The presented report deals with new results on deposition of combined coatings using Al metallization (by a plasma jet) and micro-arc (discharge) Al oxidation. After this, the coating was implanted by Ti ions with 5·10 I7 cm -2 dose (60 and 90 kV and about 200 μs duration). One series of samples with such coatings was irradiated using the accelerator Y-112 by an electron beam in melting regime (two regimes). Analysis of the structure and element composition was performed using SIMS, RBS, SEM with micro-analysis (WDS), XRD as well as measurements of microhardness, wear and adhesion. It had been demonstrated that the coating was able to sustain very high temperatures and oxidation medium. However, after electron beam irradiation temperature resistance decreased because the oxide coating was melted almost to the graphite surface. The work was funded by the Project of NANU 'Nanosystems, nanomaterials and nanotechnology'

  8. Initial Comparison of Baseline Physical and Mechanical Properties for the VHTR Candidate Graphite Grades

    Energy Technology Data Exchange (ETDEWEB)

    Carroll, Mark C. [Idaho National Lab. (INL), Idaho Falls, ID (United States). VHTR Program

    2014-09-01

    High-purity graphite is the core structural material of choice in the Very High Temperature Reactor (VHTR) design, a graphite-moderated, helium-cooled configuration capable of producing thermal energy for power generation as well as process heat for industrial applications that require temperatures higher than the outlet temperatures of present nuclear reactors. The Baseline Graphite Characterization Program is establishing accurate as-manufactured mechanical and physical property distributions in nuclear-grade graphites by providing comprehensive data that captures the level of variation in measured values. In addition to providing a thorough comparison between these values in different graphite grades, the program is also carefully tracking individual specimen source, position, and orientation information in order to provide comparisons both in specific properties and in the associated variability between different lots, different billets, and different positions from within a single billet. This report is a preliminary comparison between each of the grades of graphite that are considered “candidate” grades from four major international graphite producers. These particular grades (NBG-18, NBG-17, PCEA, IG-110, and 2114) are the major focus of the evaluations presently underway on irradiated graphite properties through the series of Advanced Graphite Creep (AGC) experiments. NBG-18, a medium-grain pitch coke graphite from SGL from which billets are formed via vibration molding, was the favored structural material in the pebble-bed configuration. NBG-17 graphite from SGL is essentially NBG-18 with the grain size reduced by a factor of two. PCEA, petroleum coke graphite from GrafTech with a similar grain size to NBG-17, is formed via an extrusion process and was initially considered the favored grade for the prismatic layout. IG-110 and 2114, from Toyo Tanso and Mersen (formerly Carbone Lorraine), respectively, are fine-grain grades produced via an isomolding

  9. The mechanical behavior and reliability prediction of the HTR graphite component at various temperature and neutron dose ranges

    International Nuclear Information System (INIS)

    Fang, Xiang; Yu, Suyuan; Wang, Haitao; Li, Chenfeng

    2014-01-01

    Highlights: • The mechanical behavior of graphite component in HTRs under high temperature and neutron irradiation conditions is simulated. • The computational process of mechanical analysis is introduced. • Deformation, stresses and failure probability of the graphite component are obtained and discussed. • Various temperature and neutron dose ranges are selected in order to investigate the effect of in-core conditions on the results. - Abstract: In a pebble-bed high temperature gas-cooled reactor (HTR), nuclear graphite serves as the main structural material of the side reflectors. The reactor core is made up of a large number of graphite bricks. In the normal operation case of the reactor, the maximum temperature of the helium coolant commonly reaches about 750 °C. After around 30 years’ full power operation, the peak value of in-core fast neutron cumulative dose reaches to 1 × 10 22 n cm −2 (EDN). Such high temperature and neutron irradiation strongly impact the behavior of graphite component, causing obvious deformation. The temperature and neutron dose are unevenly distributed inside a graphite brick, resulting in stress concentrations. The deformation and stress concentration can both greatly affect safety and reliability of the graphite component. In addition, most of the graphite properties (such as Young's modulus and coefficient of thermal expansion) change remarkably under high temperature and neutron irradiations. The irradiation-induced creep also plays a very important role during the whole process, and provides a significant impact on the stress accumulation. In order to simulate the behavior of graphite component under various in-core conditions, all of the above factors must be considered carefully. In this paper, the deformation, stress distribution and failure probability of a side graphite component are studied at various temperature points and neutron dose levels. 400 °C, 500 °C, 600 °C and 750 °C are selected as the

  10. Synthesis of carbon-13 labelled carbonaceous deposits and their evaluation for potential use as surrogates to better understand the behaviour of the carbon-14-containing deposit present in irradiated PGA graphite

    Energy Technology Data Exchange (ETDEWEB)

    Payne, L., E-mail: liam.payne@bristol.ac.uk [Interface Analysis Centre, HH Wills Physics Laboratory, University of Bristol, BS8 1TL (United Kingdom); Walker, S.; Bond, G. [Centre for Materials Science, University of Central Lancashire, PR1 2HE (United Kingdom); Eccles, H. [John Tyndall Institute for Nuclear Research, School of Computing, Engineering and Physical Sciences, University of Central Lancashire, PR1 2HE (United Kingdom); Heard, P.J.; Scott, T.B. [Interface Analysis Centre, HH Wills Physics Laboratory, University of Bristol, BS8 1TL (United Kingdom); Williams, S.J. [Radioactive Waste Management, B587, Curie Avenue, Harwell Oxford, Didcot, OX11 0RH (United Kingdom)

    2016-03-15

    The present work has used microwave plasma chemical vapour deposition to generate suitable isotopically labelled carbonaceous deposits on the surface of Pile Grade A graphite for use as surrogates for studying the behaviour of the deposits observed on irradiated graphite extracted from UK Magnox reactors. These deposits have been shown elsewhere to contain an enhanced concentration of {sup 14}C compared to the bulk graphite. A combination of Raman spectroscopy, ion beam milling with scanning electron microscopy and secondary ion mass spectrometry were used to determine topography and internal morphology in the formed deposits. Direct comparison was made against deposits found on irradiated graphite samples trepanned from a Magnox reactor core and showed a good similarity in appearance. This work suggests that the microwave plasma chemical vapour deposition technique is of value in producing simulant carbon deposits, being of sufficiently representative morphology for use in non-radioactive surrogate studies of post-disposal behaviour of {sup 14}C-containing deposits on some irradiated Magnox reactor graphite.

  11. Calculation of thermal stresses in graphite fuel blocks

    International Nuclear Information System (INIS)

    Lejeail, Y.; Cabrillat, M.T.

    2005-01-01

    This paper presents a parametric study of temperature and thermal stress calculations inside a HTGR core graphite block, taking into account the effect of fluence on the thermal and mechanical properties, up to 4. 10 21 n/cm 2 . The Finite Element model, realized with Cast3M CEA code, includes the effects of irradiation creep, which tends to produce secondary stress relaxation. Then, the Weibull weakest link theory is recalled, evaluating the possible effects of volume, stress field distribution (loading factor), and multiaxiality for graphite-type materials, and giving the methodology to compare the stress to rupture for the structure to the one obtained from characterization, in the general case. The maximum of the Weibull stress in Finite Element calculations is compared to the value for tensile specimens. It is found that the maximum of the stress corresponds to the end of the irradiation cycle, after reactor shutdown, since both thermal conductivity and Young's modulus increase with time. However, this behaviour is partly counterbalanced by the increase of material strength with irradiation. (authors)

  12. Data Report on the Newest Batch of PCEA Graphite for the VHTR Baseline Graphite Characterization Program

    Energy Technology Data Exchange (ETDEWEB)

    Carroll, Mark Christopher [Idaho National Lab. (INL), Idaho Falls, ID (United States); Cottle, David Lynn [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rohrbaugh, David Thomas [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-08-01

    This report details a comparison of mechanical and physical properties from the first billet of extruded PCEA nuclear-grade graphite from the newest batch of PCEA procured from GrafTech. Testing has largely been completed on three of the billets from the original batch of PCEA, with data distributions for those billets exhibiting a much wider range of values when compared to the distributions of properties from other grades. A higher propensity for extremely low values or specimens that broke while machining or handling was also characteristic of the billets from the first batch, owing to unusually large fissures or disparate flaws in the billets in an as-manufactured state. Coordination with GrafTech prior to placing the order for a second batch of PCEA included discussions on these large disparate flaws and how to prevent them during the manufacturing process. This report provides a comparison of the observed data distributions from properties measured in the first billet from the new batch of PCEA with those observed in the original batch, in order that an evaluation of tighter control of the manufacturing process and the outcome of these controls on final properties can be ascertained. Additionally, this billet of PCEA is the first billet to formally include measurements from two alternate test techniques that will become part of the Baseline Graphite Characterization database – the three-point bend test on sub-sized cylinders and the Brazilian disc splitting tensile strength test. As the program moves forward, property distributions from these two tests will be based on specimen geometries that match specimen geometries being used in the irradiated Advanced Graphite Creep (AGC) program. This will allow a more thorough evaluation of both the utility of the test and expected variability in properties when using those approaches on the constrained geometries of specimens irradiated in the Advanced Test Reactor as part of the AGC experiment.

  13. Graphite surveillance in N Reactor

    International Nuclear Information System (INIS)

    Woodruff, E.M.

    1991-09-01

    Graphite dimensional changes in N Reactor during its 24 yr operating history are reviewed. Test irradiation results, block measurements, stack profiles, top of reflector motion monitors, and visual observations of distortion are described. 18 refs., 14 figs., 1 tab

  14. Characterisation of Chlorine Behavior in French Graphite

    International Nuclear Information System (INIS)

    Blondel, A.; Moncoffre, N.; Toulhoat, N.; Bererd, N.; Petit, L.; Laurent, G.; Lamouroux, C.

    2016-01-01

    Chlorine 36 is one of the main radionuclides of concern for French graphite waste disposal. In order to help the understanding of its leaching behaviour under disposal conditions, the respective impact of temperature, irradiation and gas radiolysis on chlorine release in reactor has been studied. Chlorine 36 has been simulated through chlorine 37 ion implantation in virgin nuclear graphite samples. Results show that part of chlorine is highly mobile in graphite in the range of French reactors operating temperatures in relation with graphite structural recovering. Ballistic damage generated by irradiation also promotes chlorine release whereas no clear impact of the coolant gas radiolysis was observed in the absence of graphite radiolytic corrosion. (author)

  15. Nuclear graphite for high temperature reactors

    International Nuclear Information System (INIS)

    Marsden, B.J.

    2001-01-01

    The cores and reflectors in modern High Temperature Gas Cooled Reactors (HTRs) are constructed from graphite components. There are two main designs; the Pebble Bed design and the Prism design. In both of these designs the graphite not only acts as a moderator, but is also a major structural component that may provide channels for the fuel and coolant gas, channels for control and safety shut off devices and provide thermal and neutron shielding. In addition, graphite components may act as a heat sink or conduction path during reactor trips and transients. During reactor operation, many of the graphite component physical properties are significantly changed by irradiation. These changes lead to the generation of significant internal shrinkage stresses and thermal shut down stresses that could lead to component failure. In addition, if the graphite is irradiated to a very high irradiation dose, irradiation swelling can lead to a rapid reduction in modulus and strength, making the component friable.The irradiation behaviour of graphite is strongly dependent on its virgin microstructure, which is determined by the manufacturing route. Nevertheless, there are available, irradiation data on many obsolete graphites of known microstructures. There is also a well-developed physical understanding of the process of irradiation damage in graphite. This paper proposes a specification for graphite suitable for modern HTRs. (author)

  16. Graphite selection for the FMIT test cell

    International Nuclear Information System (INIS)

    Morgan, W.C.

    1982-06-01

    This document provides the basis for procuring a grade of graphite, at minimum cost, having minimum dimensional changes at low irradiation temperatures (nominal range 90 to 140 0 C). In light of those constraints, the author concludes that the most feasible approach is to attempt to reproduce a grade of graphite (TSGBF) which has exhibited a high degree of dimensional stability during low-temperature irradiations and on which irradiation-induced changes in other physical properties have been measured. The effects of differences in raw materials, especially coke morphology, and processing conditions, primarily graphitization temperture are briefly reviewed in terms of the practicality of producing a new grade of graphite with physical properties and irradiation-induced changes which would be very similar to those of TSGBF graphite. The production history and physical properties of TSGBF are also reviewed; no attempt is made, to project changes in dimensions or physical properties under the projected irradiation conditions

  17. Study by electronic microscopy of corrosion features of graphite after hot oxidation (air, 620 C)

    International Nuclear Information System (INIS)

    Jodon de Villeroche, Suzanne

    1968-01-01

    The author reports the study of corrosion features of graphite after hot oxidation in the air at 620 C. It is based on observations made by electronic microscopy. This study comes after another one dedicated to oxidation features obtained by hot corrosion of natural graphite, and aims at comparing pyrolytic graphite before and after irradiation in an atomic pile, and at performing tests on a graphite processed with ozone. After a recall of generalities about natural graphite and of some issues related to hot corrosion of natural graphite, the author presents some characteristics and features of irradiated and non-irradiated pyrolytic graphite. He reports the study of the oxidation of samples of pyrolytic graphite: production of thin lamellae, production of glaze-carbon replicates, oxidation of irradiated and of non-irradiated graphite, healing of irradiation defects, and oxidation of ozone-processed natural graphite [fr

  18. Management of UKAEA graphite liabilities

    International Nuclear Information System (INIS)

    Wise, M.

    2001-01-01

    The UK Atomic Energy Authority (UKAEA) is responsible for managing its liabilities for redundant research reactors and other active facilities concerned with the development of the UK nuclear technology programme since 1947. These liabilities include irradiated graphite from a variety of different sources including low irradiation temperature reactor graphite (the Windscale Piles 1 and 2, British Energy Pile O and Graphite Low Energy Experimental Pile at Harwell and the Material Testing Reactors at Harwell and Dounreay), advanced gas-cooled reactor graphite (from the Windscale Advanced Gas-cooled Reactor) and graphite from fast reactor systems (neutron shield graphite from the Dounreay Prototype Fast Reactor and Dounreay Fast Reactor). The decommissioning and dismantling of these facilities will give rise to over 6,000 tonnes of graphite requiring disposal. The first graphite will be retrieved from the dismantling of Windscale Pile 1 and the Windscale Advanced Gas-cooled Reactor during the next five years. UKAEA has undertaken extensive studies to consider the best practicable options for disposing of these graphite liabilities in a manner that is safe whilst minimising the associated costs and technical risks. These options include (but are not limited to), disposal as Low Level Waste, incineration, or encapsulation and disposal as Intermediate Level Waste. There are a number of technical issues associated with each of these proposed disposal options; these include Wigner energy, radionuclide inventory determination, encapsulation of graphite dust, galvanic coupling interactions enhancing the corrosion of mild steel and public acceptability. UKAEA is currently developing packaging concepts and designing packaging plants for processing these graphite wastes in consultation with other holders of graphite wastes throughout Europe. 'Letters of Comfort' have been sought from both the Low Level Waste and the Intermediate Level Waste disposal organisations to support the

  19. Sputtering characteristics of B4C-overlaid graphite for keV energy deuterium ion irradiation

    International Nuclear Information System (INIS)

    Gotoh, Y.; Yamaki, T.; Ando, T.; Jimbou, R.; Ogiwara, N.; Saidoh, M.; Teruyama, K.

    1992-01-01

    Two types of B 4 C-overlaid graphite (CFC), conversion and CVD B 4 C, together with bare CFC (PCC-2S) and/or HP B 4 C, were investigated with respect to erosion yields for 1 keV D + , D 2 /CD 4 TDS after 1 keV D + implantation, and thermal diffusivity/conductivity, in a temperature range from 300 to 1400 K. The erosion yields of both conversion and CVD B 4 C were found to be much lower than that of the bare CFC (PCC-2S), in both chemical sputtering (600-1100 K) and RES (1200-1400 K) temperature regions. The D 2 TDS peak of the conversion B 4 C was found to be located at nearly 200 K lower temperature than that of the bare CFC (PCC-2S), indicating much lower activation energy for detrapping/recombination of trapped D in the conversion B 4 C and in the CFC. The CD 4 TDS peak of the conversion B 4 C was found to be much weaker in intensity than that of the bare CFC (PCC-2S), in agreement with the present erosion yield results. Thermal diffusivities and conductivities of both the conversion B 4 C/PCC-2S and the CVD B 4 C, were measured to be nearly 1/10 of that of the bare CFC (PCC-2S), and to decrease with increasing temperatures. (orig.)

  20. Creep of Li2O

    International Nuclear Information System (INIS)

    Hollenberg, G.W.; Arthur, B.; Lui, Y.Y.

    1985-01-01

    The objective of this effort was to obtain data on the performance of lithium ceramic materials during fast neutron irradiation in support of solid breeder blanket designs. Li 2 O has been observed to swell (greater than or equal to 4%) under fast reactor irradiation. Fortunately, Li 2 O deforms at low temperatures so that swelling strains may be internally accommodated. Laboratory creep experiments were conducted between 500 to 700 0 C in order to provide data for structural analysis of in-reactor experiments and blanket design studies. A densification model agreed with most of the available data

  1. Non-covalent doping of graphitic carbon nitride with ultrathin graphene oxide and molybdenum disulfide nanosheets: an effective binary heterojunction photocatalyst under visible light irradiation.

    Science.gov (United States)

    Hu, S W; Yang, L W; Tian, Y; Wei, X L; Ding, J W; Zhong, J X; Chu, Paul K

    2014-10-01

    A proof of concept integrating binary p-n heterojunctions into a semiconductor hybrid photocatalyst is demonstrated by non-covalent doping of graphite-like carbon nitride (g-C3N4) with ultrathin GO and MoS2 nanosheets using a facile sonochemical method. In this unique ternary hybrid, the layered MoS2 and GO nanosheets with a large surface area enhance light absorption to generate more photoelectrons. On account of the coupling between MoS2 and GO with g-C3N4, the ternary hybrid possesses binary p-n heterojunctions at the g-C3N4/MoS2 and g-C3N4/GO interfaces. The space charge layers created by the p-n heterojunctions not only enhance photogeneration, but also promote charge separation and transfer of electron-hole pairs. In addition, the ultrathin MoS2 and GO with high mobility act as electron mediators to facilitate separation of photogenerated electron-hole pairs at each p-n heterojunction. As a result, the ternary hybrid photocatalyst exhibits improved photoelectrochemical and photocatalytic activity under visible light irradiation compared to other reference materials. The results provide new insights into the large-scale production of semiconductor photocatalysts. Copyright © 2014 Elsevier Inc. All rights reserved.

  2. Artificial graphites

    International Nuclear Information System (INIS)

    Maire, J.

    1984-01-01

    Artificial graphites are obtained by agglomeration of carbon powders with an organic binder, then by carbonisation at 1000 0 C and graphitization at 2800 0 C. After description of the processes and products, we show how the properties of the various materials lead to the various uses. Using graphite enables us to solve some problems, but it is not sufficient to satisfy all the need of the application. New carbonaceous material open application range. Finally, if some products are becoming obsolete, other ones are being developed in new applications [fr

  3. Simultaneous consolidation and creep

    DEFF Research Database (Denmark)

    Krogsbøll, Anette

    1997-01-01

    Materials that exhibit creep under constant effective stress typically also show rate dependent behavior. The creep deformations and the rate sensitive behavior is very important when engineering and geological problems with large time scales are considered. When stress induced compaction...

  4. Creep of crystals

    International Nuclear Information System (INIS)

    Poirier, J.-P.

    1988-01-01

    Creep mechanisms for metals, ceramics and rocks, effect of pressure and temperature on deformation processes are considered. The role of crystal defects is analysed, different models of creep are described. Deformation mechanisms maps for different materials are presented

  5. AGC-2 Irradiation Data Qualification Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Laurence C. Hull

    2012-07-01

    The Graphite Technology Development Program will run a series of six experiments to quantify the effects of irradiation on nuclear grade graphite. The second Advanced Graphite Creep (AGC) experiment (AGC-2) began with Advanced Test Reactor (ATR) Cycle 149A on April 12, 2011, and ended with ATR Cycle 151B on May 5, 2012. The purpose of this report is to qualify AGC-2 irradiation monitoring data following INL Management and Control Procedure 2691, Data Qualification. Data that are Qualified meet the requirements for data collection and use as described in the experiment planning and quality assurance documents. Data that do not meet the requirements are Failed. Some data may not quite meet the requirements, but may still provide some useable information. These data are labeled as Trend. No Trend data were identified for the AGC-2 experiment. All thermocouples functioned throughout the AGC-2 experiment. There was one instance where spurious signals or instrument power interruption resulted in a recorded temperature value being well outside physical reality. This value was identified and labeled as Failed data. All other temperature data are Qualified. All helium and argon gas flow data are within expected ranges. Total gas flow was approximately 50 sccm through the capsule. Helium gas flow was briefly increased to 100 sccm during reactor shutdown. All gas flow data are Qualified. At the start of the experiment, moisture in the outflow gas line increased to 200 ppmv then declined to less than 10 ppmv over a period of 5 days. This increase in moisture coincides with the initial heating of the experiment and drying of the system. Moisture slightly exceeded 10 ppmv three other times during the experiment. While these moisture values exceed the 10 ppmv threshold value, the reported measurements are considered accurate and to reflect moisture conditions in the capsule. All moisture data are Qualified. Graphite creep specimens are subjected to one of three loads, 393 lbf

  6. High-resolution TEM microscopy study of the creep behaviour of carbon-based cathode materials

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Wei, E-mail: wwlyzwkj@126.com [College of Materials Science and Engineering, Henan University of Science and Technology, Luoyang 471023 (China); Collaborative Innovation Center of Nonferrous Metals Henan Province, Luoyang 471023 (China); Chen, Weijie [College of Materials Science and Engineering, Henan University of Science and Technology, Luoyang 471023 (China); Collaborative Innovation Center of Nonferrous Metals Henan Province, Luoyang 471023 (China); Gu, Wanduo [Collaborative Innovation Center of Nonferrous Metals Henan Province, Luoyang 471023 (China)

    2017-02-27

    Creep is in close relationship with the materials deterioration and deformation of the cathodes in aluminum reduction cells. The purpose of this work is to obtain the creep mechanism of the carbon cathode for aluminum electrolysis. A modified Rapoport equipment was used for measuring the creep strain of the semi-graphitic cathodes during aluminum electrolysis with CR=2.5 and at temperature of 945 ℃. The arrangement of carbon atom has been studied after hexagonal graphite converting into rhombohedral graphite during aluminum electrolysis by XRD and high-resolution transmission electron microscopy (HRTEM). The creep deformation of the carbon cathode has a close relationship with the mobile dislocation walls. These results will be helpful in controlling the cathode quality and its performance in aluminum reduction cells.

  7. Creep in ceramics

    CERN Document Server

    Pelleg, Joshua

    2017-01-01

    This textbook is one of its kind, since there are no other books on Creep in Ceramics. The book consist of two parts: A and B. In part A general knowledge of creep in ceramics is considered, while part B specifies creep in technologically important ceramics. Part B covers creep in oxide ceramics, carnides and nitrides. While covering all relevant information regarding raw materials and characterization of creep in ceramics, the book also summarizes most recent innovations and developments in this field as a result of extensive literature search.

  8. High flux irradiations of Li coatings on polycrystalline W and ATJ graphite with D, He, and He-seeded D plasmas at Magnum PSI

    NARCIS (Netherlands)

    Neff, A. L.; Allain, J. P.; F. Bedoya,; Morgan, T. W.; De Temmerman, G.

    2015-01-01

    Lithium wall conditioning on PFCs (Plasma Facing Components) on a variety of substrate platforms (e.g. graphite, Mo, etc.) has resulted in improved plasma performance on multiple magnetic fusion devices. On graphite, this improvement occurs through the control of retention and recycling of hydrogen

  9. Methodology of characterization of radioactive graphite

    International Nuclear Information System (INIS)

    Pina, G.; Rodriguez, M.; Lara, E.; Magro, E.; Gascon, J. L.; Leganes, J. L.

    2014-01-01

    Since the dismantling of Vandellos I, ENRESA has promoted the precise knowledge of the inventory of irradiated graphite (graphite-i) through establishing methodologies for radiological characterization of the vector of radionuclides of interest and their correlations as the primary means of characterization strategy to establish the safer management of this material in its life cycle. (Author)

  10. Irradiation induced effects in zirconium (A review)

    International Nuclear Information System (INIS)

    Madden, P.K.

    1975-06-01

    Irradiation creep in zirconium and its alloys is comprehensively discussed. The main theories are outlined and the gaps between them and the observed creep behaviour, indicated. Although irradiation induced point defects play an important role, effects due to irradiation induced dislocation loops seem insignificant. The experimental results suggest that microstructural variations due to prior cold-working or hydrogen injection perturb the irradiation growth and the irradiation creep of zircaloy. Further investigations into these areas are required. One disadvantage of creep experiments lies in their duration. The possibility of accelerated experiments using ion implantation or electron irradiation is examined in the final section, and its possible advantages and disadvantages are outlined. (author)

  11. In situ monitored in-pile creep testing of zirconium alloys

    Science.gov (United States)

    Kozar, R. W.; Jaworski, A. W.; Webb, T. W.; Smith, R. W.

    2014-01-01

    The experiments described herein were designed to investigate the detailed irradiation creep behavior of zirconium based alloys in the HALDEN Reactor spectrum. The HALDEN Test Reactor has the unique capability to control both applied stress and temperature independently and externally for each specimen while the specimen is in-reactor and under fast neutron flux. The ability to monitor in situ the creep rates following a stress and temperature change made possible the characterization of creep behavior over a wide stress-strain-rate-temperature design space for two model experimental heats, Zircaloy-2 and Zircaloy-2 + 1 wt%Nb, with only 12 test specimens in a 100-day in-pile creep test program. Zircaloy-2 specimens with and without 1 wt% Nb additions were tested at irradiation temperatures of 561 K and 616 K and stresses ranging from 69 MPa to 455 MPa. Various steady state creep models were evaluated against the experimental results. The irradiation creep model proposed by Nichols that separates creep behavior into low, intermediate, and high stress regimes was the best model for predicting steady-state creep rates. Dislocation-based primary creep, rather than diffusion-based transient irradiation creep, was identified as the mechanism controlling deformation during the transitional period of evolving creep rate following a step change to different test conditions.

  12. Creep fatigue assessment for EUROFER components

    Energy Technology Data Exchange (ETDEWEB)

    Özkan, Furkan, E-mail: oezkan.furkan@partner.kit.edu; Aktaa, Jarir

    2015-11-15

    Highlights: • Design rules for creep fatigue assessment are developed to EUROFER components. • Creep fatigue assessment tool is developed in FORTRAN code with coupling MAPDL. • Durability of the HCPB-TBM design is discussed under typical fusion reactor loads. - Abstract: Creep-fatigue of test blanket module (TBM) components built from EUROFER is evaluated based on the elastic analysis approach in ASME Boiler Pressure Vessel Code (BPVC). The required allowable number of cycles design fatigue curve and stress-to-rupture curve to estimate the creep-fatigue damage are used from the literature. Local stress, strain and temperature inputs for the analysis of creep-fatigue damage are delivered by the finite element code ANSYS utilizing the Mechanical ANSYS Parametric Design Language (MAPDL). A developed external FORTRAN code used as a post processor is coupled with MAPDL. Influences of different pulse durations (hold-times) and irradiation on creep-fatigue damage for the preliminary design of the Helium Cooled Pebble Bed Test Blanket Module (HCPB-TBM) are discussed for the First Wall component of the TBM box.

  13. Effects of composition on the in-reactor creep of AISI 316

    International Nuclear Information System (INIS)

    Bates, J.F.; Gilbert, E.R.

    1979-08-01

    In-reactor tests designed to provide information on the relationship between compositional variations and irradiation-induced swelling and creep have achieved an exposure of 4.6 x 10 22 n/cm 2 (E > 0.1 MeV) at 450 0 C. Postirradiation diametral measurements of pressurized tube specimens have indicated that irradiation-induced creep of 316 stainless steel can be modified by compositional variations of minor alloying elements. There is a general trend for specimens with higher swelling to exhibit higher creep. Silicon, phosphorus and molybdenum all retard in-reactor creep and inhibit irradiation-induced swelling as well. However, the relationship between creep and swelling is strongly composition dependent. The data suggest that carbon and nitrogen act synergistically the major influence being the nitrogen concentration. The irradiation-induced creep is insensitive to cobalt variations to the fluences investigated

  14. Effects of composition on the in-reactor creep of AISI 316

    International Nuclear Information System (INIS)

    Bates, J.F.; Gilbert, E.R.

    1980-01-01

    Pre- and postirradiation measurements of pressurized tube specimens irradiated at 450/degree/C to 4.6*10/sup 22/ n/cm/sup 2/(E>0.1 MeV) have indicated that increases in the solute concentrations of silicon, phosphorus, and molybdenum retard irradiation creep. The data suggest that carbon and nitrogen act synergistically with the major influence on creep being the nitrogen concentration. Irradiation-induced creep is insensitive to cobalt variations. There is a trend for specimens with higher swelling to exhibit higher creep. As the shear modulus increases, irradiation creep also increases. This shear modulus correlation is opposite to one observed for thermal creep deformation. 8 refs

  15. Creep of Li2O

    International Nuclear Information System (INIS)

    Hollenberg, G.W.; Liu, Y.Y.; Arthur, B.

    1984-11-01

    The tritium breeding material with the highest lithium atom density, Li 2 O has been observed to incur significant swelling (>4%) under fast reactor irradiation. Such swelling, if unrestrained leads to either unacceptable, induced-strains in adjacent structural material or undesirable design compromises. Fortunately, however, Li 2 O deforms at low temperatures so that swelling strains may be internally accommodated. Laboratory dilational creep experiments were conducted on unirradiated Li 2 O between 500 and 700 0 C in order to provide data for structural analysis of in-reactor experiments and blanket design studies. A densification model agreed with most of the available data

  16. A graphite foam reinforced by graphite particles

    Energy Technology Data Exchange (ETDEWEB)

    Zhu, J.J.; Wang, X.Y.; Guo, L.F.; Wang, Y.M.; Wang, Y.P.; Yu, M.F.; Lau, K.T.T. [DongHua University, Shanghai (China). College of Material Science and Engineering

    2007-11-15

    Graphite foam was obtained after carbonization and graphitization of a pitch foam formed by the pyrolysis of coal tar based mesophase pitch mixed with graphite particles in a high pressure and temperature chamber. The graphite foam possessed high mechanical strength and exceptional thermal conductivity after adding the graphite particles. Experimental results showed that the thermal conductivity of modified graphite foam reached 110W/m K, and its compressive strength increased from 3.7 MPa to 12.5 MPa with the addition of 5 wt% graphite particles. Through the microscopic observation, it was also found that fewer micro-cracks were formed in the cell wall of the modified foam as compared with pure graphite foam. The graphitization degree of modified foam reached 84.9% and the ligament of graphite foam exhibited high alignment after carbonization at 1200{sup o}C for 3 h and graphitization at 3000{sup o}C for 10 min.

  17. Nanoindentation creep versus bulk compressive creep of dental resin-composites.

    Science.gov (United States)

    El-Safty, S; Silikas, N; Akhtar, R; Watts, D C

    2012-11-01

    To evaluate nanoindentation as an experimental tool for characterizing the viscoelastic time-dependent creep of resin-composites and to compare the resulting parameters with those obtained by bulk compressive creep. Ten dental resin-composites: five conventional, three bulk-fill and two flowable were investigated using both nanoindentation creep and bulk compressive creep methods. For nano creep, disc specimens (15mm×2mm) were prepared from each material by first injecting the resin-composite paste into metallic molds. Specimens were irradiated from top and bottom surfaces in multiple overlapping points to ensure optimal polymerization using a visible light curing unit with output irradiance of 650mW/cm(2). Specimens then were mounted in 3cm diameter phenolic ring forms and embedded in a self-curing polystyrene resin. Following grinding and polishing, specimens were stored in distilled water at 37°C for 24h. Using an Agilent Technologies XP nanoindenter equipped with a Berkovich diamond tip (100nm radius), the nano creep was measured at a maximum load of 10mN and the creep recovery was determined when each specimen was unloaded to 1mN. For bulk compressive creep, stainless steel split molds (4mm×6mm) were used to prepare cylindrical specimens which were thoroughly irradiated at 650mW/cm(2) from multiple directions and stored in distilled water at 37°C for 24h. Specimens were loaded (20MPa) for 2h and unloaded for 2h. One-way ANOVA, Levene's test for homogeneity of variance and the Bonferroni post hoc test (all at p≤0.05), plus regression plots, were used for statistical analysis. Dependent on the type of resin-composite material and the loading/unloading parameters, nanoindentation creep ranged from 29.58nm to 90.99nm and permanent set ranged from 8.96nm to 30.65nm. Bulk compressive creep ranged from 0.47% to 1.24% and permanent set ranged from 0.09% to 0.38%. There was a significant (p=0.001) strong positive non-linear correlation (r(2)=0.97) between bulk

  18. Experience with graphite in JET

    International Nuclear Information System (INIS)

    Pick, M.A.; Celentano, G.; Deksnis, E.; Dietz, K.J.; Shaw, R.; Sonnenberg, K.; Walravens, M.

    1987-01-01

    During the current operational period of JET more than 50% of the internal area of the machine is covered in graphite tiles. This includes the 15 m 2 of carbon tiles installed in the new toroidal limiter, the 40 poloidal belts of graphite tiles covering the U-joints and bellows as well as a two metre high ring (-- 20 m 2 ) or carbon tiles on the inner wall of the Torus. A ring of tiles in the equatorial plane (3 tiles high) consists of carbon-carbon fibre tiles. Test bed results indicated that the fine grained graphite tiles cracked at ∼ 1 kW/cm 2 for 2s of irradiation whereas the carbon-carbon fibre tiles were able to sustain a flux, limited by the irradiation facility, of 3.5 kW for 3s without any damage. The authors report on the generally positive experience they have had had with the installed graphite during the present and previous in-vessel configurations. This includes the physical integrity of the tiles under severe conditions such as high energy run-away electron beams, plasma disruptions and high heat fluxes. They report on the importance of the precise positioning of the inner wall and x-point tiles at the very high power fluxes of JET and the effect of deviations on both graphite and carbon-fibre tiles

  19. Creep fatigue design of FBR components

    International Nuclear Information System (INIS)

    Bhoje, S.B.; Chellapandi, P.

    1997-01-01

    This paper deals with the characteristic features of Fast Breeder Reactor (FBR) with reference to creep fatigue, current creep fatigue design approach in compliance with RCCMR (1987) design code, material data, effects of weldments and neutron irradiation, material constitutive models employed, structural analysis and further R and D required for achieving maturity in creep fatigue design of FBR components. For the analysis reported in this paper, material constitutive models developed based on ORNIb (Oak Ridge National Laboratory) and Chaboche viscoplastic theories are employed to demonstrate the potential of FBR components for higher plant temperatures and/or longer life. The results are presented for the studies carried out towards life prediction of Prototype Fast Breeder Reactor (PFBR) components. (author). 24 refs, 8 figs, 5 tabs

  20. Progress in radioactive graphite waste management

    International Nuclear Information System (INIS)

    2010-07-01

    , especially in the UK. It is intended that this report which contains the proceedings of the conference should contribute to progress in the management of radioactive graphite worldwide. The report contains a selection of the papers presented on various issues related to dismantling and treating irradiated graphite. In addition, the report contains summaries of the four topical discussions which were held during the conference

  1. Creep in buffer clay

    International Nuclear Information System (INIS)

    Pusch, R.; Adey, R.

    1999-12-01

    The study involved characterization of the microstructural arrangement and molecular forcefields in the buffer clay for getting a basis for selecting suitable creep models. It is concluded that the number of particles and wide range of the particle bond spectrum require that stochastical mechanics and thermodynamics will be considered and they are basic to the creep model proposed for predicting creep settlement of the canisters. The influence of the stress level on creep strain of MX-80 clay is not well known but for the buffer creep is approximately proportional to stress. Theoretical considerations suggest a moderate impact for temperatures up to 90 deg C and this is supported by model experiments. It is believed that the assumption of strain being proportional to temperature is conservative. The general performance of the stochastic model can be illustrated in principle by use of visco-elastic rheological models implying a time-related increase in viscosity. The shear-induced creep settlement under constant volume conditions calculated by using the proposed creep model is on the order of 1 mm in ten thousand years and up to a couple of millimeters in one million years. It is much smaller than the consolidation settlement, which is believed to be on the order of 10 mm. The general conclusion is that creep settlement of the canisters is very small and of no significance to the integrity of the buffer itself or of the canisters

  2. Thermally exfoliated graphite oxide

    Science.gov (United States)

    Prud'Homme, Robert K. (Inventor); Aksay, Ilhan A. (Inventor); Abdala, Ahmed (Inventor)

    2011-01-01

    A modified graphite oxide material contains a thermally exfoliated graphite oxide with a surface area of from about 300 sq m/g to 2600 sq m/g, wherein the thermally exfoliated graphite oxide displays no signature of the original graphite and/or graphite oxide, as determined by X-ray diffraction.

  3. Creep buckling of shells

    International Nuclear Information System (INIS)

    Stone, C.M.; Nickell, R.E.

    1977-01-01

    Because of the characteristics of LMFBR primary piping components (thin-walled, low pressure, high temperature), the designer must guard against creep buckling as a potential failure mode for certain critical regions, such as elbows, where structural flexibility and inelastic response may combine to concentrate deformation and cause instability. The ASME Boiler and Pressure Vessel Code, through its elevated temperature Code Case 1592 (Section III, Division 1) provides design rules for Class 1 components aimed at preventing creep buckling during the design life. A similar set of rules is being developed for Class 2 and 3 components at this time. One of the original concepts behind the creep buckling rules was that the variability in creep properties (especially due to the effects of prior heat treatment), the uncertainty about initial imperfections, and the lack of confirmed accuracy of design analysis meant that conservatism would be difficult to assure. As a result, a factor of ten on service life was required (i.e. analysis must show that, under service conditions that extrapolate the life of the component by ten times, creep buckling does not occur). Two obvious problems with this approach are that: first, the creep behavior must also be extrapolated (since most creep experiments are terminated at a small fraction of the design life, extrapolation of creep data is already an issue, irrespective of the creep buckling question); second the nonlinear creep analysis, which is very nearly prohibitively expensive for design life histograms, becomes even more costly. Analytical results for an aluminum cylindrical shell subjected to axial loads at elevated temperatures are used to examine the supposed equivalence of two types of time-dependent buckling safety factors - a factor of ten on service life and a factor of 1.5 on loading

  4. Spherical Indentation Techniques for Creep Property Evaluation Considering Transient Creep

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Dongkyu; Kim, Minsoo; Lee, Hyungyil [Sogang Univ., Seoul, (Korea, Republic of); Lee, Jin Haeng [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-11-15

    Creep through nanoindentations has attracted increasing research attention in recent years. Many studies related to indentation creep tests, however, have simply focused on the characteristics of steady-state creep, and there exist wide discrepancies between the uniaxial test and the indentation test. In this study, we performed a computational simulation of spherical indentations, and we proposed a method for evaluating the creep properties onsidering transient creep. We investigated the material behavior with variation of creep properties and expressed it using regression equations for normalized variables. We finally developed a program to evaluate the creep properties considering transient creep. By using the proposed method, we successfully obtained creep exponents with an average error less than 1.1 and creep coefficients with an average error less than 2.3 from the load-depth curve.

  5. Spherical Indentation Techniques for Creep Property Evaluation Considering Transient Creep

    International Nuclear Information System (INIS)

    Lim, Dongkyu; Kim, Minsoo; Lee, Hyungyil; Lee, Jin Haeng

    2013-01-01

    Creep through nanoindentations has attracted increasing research attention in recent years. Many studies related to indentation creep tests, however, have simply focused on the characteristics of steady-state creep, and there exist wide discrepancies between the uniaxial test and the indentation test. In this study, we performed a computational simulation of spherical indentations, and we proposed a method for evaluating the creep properties onsidering transient creep. We investigated the material behavior with variation of creep properties and expressed it using regression equations for normalized variables. We finally developed a program to evaluate the creep properties considering transient creep. By using the proposed method, we successfully obtained creep exponents with an average error less than 1.1 and creep coefficients with an average error less than 2.3 from the load-depth curve

  6. Creep properties of discontinuous fibre composites with partly creeping fibres

    International Nuclear Information System (INIS)

    Bilde-Soerensen, J.B.; Lilholt, H.

    1977-05-01

    In a previous report (RISO-M-1810) the creep properties of discontinuous fibre composites with non-creeping fibres were analyzed. In the present report this analysis is extended to include the case of discontinuous composites with partly creeping fibres. It is shown that the creep properties of the composite at a given strain rate, epsilonsub(c), depend on the creep properties of the matrix at a strain rate higher than epsilonsub(c), and on the creep properties of the fibres at epsilonsub(c). The composite creep law is presented in a form which permits a graphical determination of the composite creep curve. This can be constructed on the basis of the matrix and the fibre creep curves by vector operations in a log epsilon vs. log sigma diagram. The matrix contribution to the creep strength can be evaluated by a simple method. (author)

  7. A Graphite Isotope Ratio Method: A Primer on Estimating Plutonium Production in Graphite Moderated Reactors

    International Nuclear Information System (INIS)

    Gesh, Christopher J.

    2004-01-01

    The Graphite Isotope Ratio Method (GIRM) is a technique used to estimate the total plutonium production in a graphite-moderated reactor. The cumulative plutonium production in that reactor can be accurately determined by measuring neutron irradiation induced isotopic ratio changes in certain impurity elements within the graphite moderator. The method does not require detailed knowledge of a reactor's operating history, although that knowledge can decrease the uncertainty of the production estimate. The basic premise of the Graphite Isotope Ratio Method is that the fluence in non-fuel core components is directly related to the cumulative plutonium production in the nuclear fuel

  8. Metallurgical principles of creep processes

    International Nuclear Information System (INIS)

    Bolton, C.J.

    1977-12-01

    A brief review is presented of current theories of a number of the physical processes which can be involved in deformation and fracture under creep conditions. The processes considered are power law creep, diffusion creep, grain boundary sliding, cavitation and other modes of failure, and creep crack growth. The note concludes with some suggestions for future work. (author)

  9. Bridged graphite oxide materials

    Science.gov (United States)

    Herrera-Alonso, Margarita (Inventor); McAllister, Michael J. (Inventor); Aksay, Ilhan A. (Inventor); Prud'homme, Robert K. (Inventor)

    2010-01-01

    Bridged graphite oxide material comprising graphite sheets bridged by at least one diamine bridging group. The bridged graphite oxide material may be incorporated in polymer composites or used in adsorption media.

  10. The optical properties and photocatalytic activity of CdS-ZnS-TiO{sub 2}/Graphite for isopropanol degradation under visible light irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Rahmawati, Fitria, E-mail: fitria@mipa.uns.ac.id; Wulandari, Rini, E-mail: riniwulandari55@yahoo.com; Murni, Irvinna M., E-mail: irvinna-mutiara@yahoo.com; Mudjijono, E-mail: mbahparto@yahoo.com [Research Group of Solid State Chemistry & Catalysis, Chemistry Department, Sebelas Maret University, Jl. Ir. Sutami 36 A Kentingan, Surakarta, 57126 (Indonesia)

    2016-02-08

    This research prepared a photocatalyst tablet of CdS-ZnS-TiO{sub 2} on a graphite substrate. The synthesis was conducted through chemical bath deposition method. The graphite substrate used was a waste graphite rod from primary batteries. The aims of this research are studying the crystal structure, the optical properties and the photocatalytic activity of the prepared material. The photocatalytic activity was determined through isopropanol degradation. The result shows that the TiO{sub 2}/Graphite provide direct transition gap energy at 2.91 eV and an indirect transition gap energy at 3.21 eV. Deposition of CdS-ZnS changed the direct transition gap energy to 3.01 eV and the indirect transition gap energy to 3.22 eV. Isopropanol degradation with the prepared catalyst produced new peaks at 223-224 nm and 265-266 nm confirming the production of acetone. The degradation follows first order with rate constant of 2.4 × 10{sup −2} min{sup −1}.

  11. Stress-affected microstructural development and creep-swelling interrelationship

    International Nuclear Information System (INIS)

    Brager, H.R.; Garner, F.A.; Gilbert, E.R.; Flinn, J.E.; Wolfer, W.G.

    1977-05-01

    Macroscopic measurement of the deformations arising from swelling and creep during neutron irradiation indicate that both processes are dependent on the magnitude and possibly the sign of the applied stress state. Current modeling efforts also indicate that a strong interaction exists between swelling and creep through the stress state. Because the macroscopic distortions arise from the integrated microscopic strains associated with specific microstructural elements, the effect of applied stress on microstructural development has been studied

  12. Microstructure Evolution During Creep of Cold Worked Austenitic Stainless Steel

    Science.gov (United States)

    Krishan Yadav, Hari; Ballal, A. R.; Thawre, M. M.; Vijayanand, V. D.

    2018-04-01

    The 14Cr–15Ni austenitic stainless steel (SS) with additions of Ti, Si, and P has been developed for their superior creep strength and better resistance to void swelling during service as nuclear fuel clad and wrapper material. Cold working induces defects such as dislocations that interact with point defects generated by neutron irradiation and facilitates recombination to make the material more resistant to void swelling. In present investigation, creep properties of the SS in mill annealed condition (CW0) and 40 % cold worked (CW4) condition were studied. D9I stainless steel was solution treated at 1333 K for 30 minutes followed by cold rolling. Uniaxial creep tests were performed at 973 K for various stress levels ranging from 175-225 MPa. CW4 samples exhibited better creep resistance as compared to CW0 samples. During creep exposure, cold worked material exhibited phenomena of recovery and recrystallization wherein new strain free grains were observed with lesser dislocation network. In contrast CW0 samples showed no signs of recovery and recrystallization after creep exposure. Partial recrystallization on creep exposure led to higher drop in hardness in cold worked sample as compared to that in mill annealed sample. Accelerated precipitation of carbides at the grain boundaries was observed during creep exposure and this phenomenon was more pronounced in cold worked sample.

  13. Process for purifying graphite

    International Nuclear Information System (INIS)

    Clausius, R.A.

    1985-01-01

    A process for purifying graphite comprising: comminuting graphite containing mineral matter to liberate at least a portion of the graphite particles from the mineral matter; mixing the comminuted graphite particles containing mineral matter with water and hydrocarbon oil to form a fluid slurry; separating a water phase containing mineral matter and a hydrocarbon oil phase containing grahite particles; and separating the graphite particles from the hydrocarbon oil to obtain graphite particles reduced in mineral matter. Depending upon the purity of the graphite desired, steps of the process can be repeated one or more times to provide a progressively purer graphite

  14. Creep-fatigue deformation behaviour of OFHC-copper and CuCrZr alloy with different heat treatments and with and without neutron irradiation

    DEFF Research Database (Denmark)

    Singh, B.N.; Li, M.; Stubbins, J.F.

    2005-01-01

    results of these investigations are presented andtheir implications are briefly discussed in the present report. The central conclusion emerging from the present work is that the application of holdtime generally reduces the number of cycles to failure. The largest reduction was found to be in the caseof...... OFHC-copper. Surprisingly, the magnitude of this reduction is found to be larger at lower levels of strain or stress amplitudes, particularly when the level of the stress amplitude is below the monotonic yield strength of the material. The reduction inthe yield strength due to overaging heat treatments...... causes a substantial decrease in the number of cycles to failure at all holdtimes investigated. The increase in the yield strength due to neutron irradiation at 333 K, on the other hand, causes anincrease in the number of cycles to failure. The irradiation at 573 K to a dose level of 0.2-0.3 dpa does...

  15. Radiation-induced creep and swelling

    International Nuclear Information System (INIS)

    Heald, P.T.

    1977-01-01

    The physical basis for radiation induced creep and swelling is reviewed. The interactions between the point defects and dislocations are recalled since these interactions are ultimately responsible for the observable deformation phenomena. Both the size misfit interaction and the induced inhomogeneity interaction are considered since the former gives rise to irradiation swelling while the latter, which depends on both internal and external stresses, results in irradiation creep. The defect kinetics leading to the deformation processes are discussed in terms of chemical rate theory. The rate equations for the spatially averaged interstitial and vacancy concentrations are expressed in terms of the microstructural sink strengths and the solution of these equations leads to general expressions for the deformation rates

  16. Seismic Creep, USA Images

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Seismic creep is the constant or periodic movement on a fault as contrasted with the sudden rupture associated with an earthquake. It is a usually slow deformation...

  17. Biaxial Creep Specimen Fabrication

    International Nuclear Information System (INIS)

    JL Bump; RF Luther

    2006-01-01

    This report documents the results of the weld development and abbreviated weld qualification efforts performed by Pacific Northwest National Laboratory (PNNL) for refractory metal and superalloy biaxial creep specimens. Biaxial creep specimens were to be assembled, electron beam welded, laser-seal welded, and pressurized at PNNL for both in-pile (JOYO reactor, O-arai, Japan) and out-of-pile creep testing. The objective of this test campaign was to evaluate the creep behavior of primary cladding and structural alloys under consideration for the Prometheus space reactor. PNNL successfully developed electron beam weld parameters for six of these materials prior to the termination of the Naval Reactors program effort to deliver a space reactor for Project Prometheus. These materials were FS-85, ASTAR-811C, T-111, Alloy 617, Haynes 230, and Nirnonic PE16. Early termination of the NR space program precluded the development of laser welding parameters for post-pressurization seal weldments

  18. Biaxial Creep Specimen Fabrication

    Energy Technology Data Exchange (ETDEWEB)

    JL Bump; RF Luther

    2006-02-09

    This report documents the results of the weld development and abbreviated weld qualification efforts performed by Pacific Northwest National Laboratory (PNNL) for refractory metal and superalloy biaxial creep specimens. Biaxial creep specimens were to be assembled, electron beam welded, laser-seal welded, and pressurized at PNNL for both in-pile (JOYO reactor, O-arai, Japan) and out-of-pile creep testing. The objective of this test campaign was to evaluate the creep behavior of primary cladding and structural alloys under consideration for the Prometheus space reactor. PNNL successfully developed electron beam weld parameters for six of these materials prior to the termination of the Naval Reactors program effort to deliver a space reactor for Project Prometheus. These materials were FS-85, ASTAR-811C, T-111, Alloy 617, Haynes 230, and Nirnonic PE16. Early termination of the NR space program precluded the development of laser welding parameters for post-pressurization seal weldments.

  19. Limit analysis via creep

    International Nuclear Information System (INIS)

    Taroco, E.; Feijoo, R.A.

    1981-07-01

    In this paper it is presented a variational method for the limit analysis of an ideal plastic solid. This method has been denominated as Modified Secundary Creep and enables to find the collapse loads through a minimization of a functional and a limit process. Given an ideal plastic material it is shown how to determinate the associated secundary creep constitutive equation. Finally, as an application, it is found the limit load in an pressurized von Mises rigid plastic sphere. (Author) [pt

  20. Evaluation of results from an in-pile creep test in the Studsvik R2 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pettersson, Kjell [Entropy Materials, Stockholm (Sweden)

    2002-01-01

    An in-pile creep test with bowing of cladding tubes has been performed in a hot water loop in the Studsvik R2 reactor . One test was performed in the core and one outside the core. The out-of-pile sample showed some minor primary creep strain while the in-pile specimen deformed at a steady rate of 5x10{sup -7}/h . However, when the results were compared to a broader data base of Zircaloy in-pile creep it became clear that the creep deformation observed is a primary creep which occurs before the irradiation creep in Zircaloy reaches a constant steady state creep rate. This primary stage is interpreted as a consequence of the development of an irradiation induced microstructure in Zircaloy which does not reach a steady state until a dose of about 10{sup 21} n/cm{sup 2} . At this stage the steady state irradiation creep starts. From this interpretation it is concluded that it is quite feasible to use the test method on pre-irradiated material in which it can be expected that the steady state will be reached already after short irradiation times.

  1. Heat Transfer During Evaporation of Cesium From Graphite Surface in an Argon Environment

    Directory of Open Access Journals (Sweden)

    Bespala Evgeny

    2016-01-01

    Full Text Available The article focuses on discussion of problem of graphite radioactive waste formation and accumulation. It is shown that irradiated nuclear graphite being inalienable part of uranium-graphite reactor may contain fission and activation products. Much attention is given to the process of formation of radioactive cesium on the graphite element surface. It is described a process of plasma decontamination of irradiated graphite in inert argon atmosphere. Quasi-one mathematical model is offered, it describes heat transfer process in graphite-cesium-argon system. Article shows results of calculation of temperature field inside the unit cell. Authors determined the factors which influence on temperature change.

  2. Contribution of dislocation creep to the radiational creep of materials

    International Nuclear Information System (INIS)

    Borodin, V.A.; Ryazanov, A.I.

    1986-01-01

    The authors propose a model of the orientational dependences of the preferences of discrete linear dislocations in which the influence of the external load on the step concentration at the dislocations is taken into account. The use of this model, taking into account the mechanism of stress-induced anisotropy of the elastic interaction between point defects and dislocations, not only permits a correct qualitative explanation of the dependence of the rate of radiational creep on the basic irradiation parameters (dose, stress, temperature) but also allows approximate quantitative agreement with experimental results to be obtained. At sufficiently high stress, the theory predicts conditions of the formation of an ensemble of dislocational loops with a specific direction of the Burgers vector

  3. Elastic properties of graphite and interstitial defects

    International Nuclear Information System (INIS)

    Ayasse, J.-B.

    1977-01-01

    The graphite elastic constants C 33 and C 44 , reflecting the interaction of the graphitic planes, were experimentally measured as a function of irradiation and temperature. A model of non-central strength atomic interaction was established to explain the experimental results obtained. This model is valid at zero temperature. The temperature dependence of the elastic properties was analyzed. The influence of the elastic property variations on the specific heat of the lattice at very low temperature was investigated [fr

  4. Creep feeding nursing beef calves.

    Science.gov (United States)

    Lardy, Gregory P; Maddock, Travis D

    2007-03-01

    Creep feeding can be used to increase calf weaning weights. However, the gain efficiency of free-choice, energy-based creep feeds is relatively poor. Generally, limit-feeding, high-protein creep feeds are more efficient, and gains may be similar to those produced by creep feeds offered free choice. Creep feeding can increase total organic matter intake and improve the overall energy status of the animal. Creep-fed calves tend to acclimate to the feedlot more smoothly than unsupplemented calves. Furthermore, provision of a high-starch creep feed may have a positive influence on subsequent carcass quality traits. Creep feeding can be applied to numerous environmental situations to maximize calf performance; however, beef cattle producers should consider their individual situations carefully before making the decision to creep feed.

  5. Thermal creep and stress-affected precipitation of 20% cold-worked 316 stainless steel

    International Nuclear Information System (INIS)

    Puigh, R.J.; Lovell, A.J.; Garner, F.A.

    1984-01-01

    Measurements of the thermal creep of 20% cold-worked 316 stainless steel have been performed for temperatures from 593 to 760 0 C, stress levels as high as 138 MPa and exposure times as long as 15,000 hours. The creep strains exhibit a complex behavior arising from the combined action of true creep and stress-affected precipitation of intermetallic phases. The latter process is suspected to be altered by neutron irradiation. (orig.)

  6. AGC-3 Experiment Irradiation Monitoring Data Qualification Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Hull, Laurence C. [Idaho National Lab. (INL), Idaho Falls, ID (United States). VHTR Technology Development Office

    2014-08-01

    The Graphite Technology Development Program will run a series of six experiments to quantify the effects of irradiation on nuclear-grade graphite. The third experiment, Advanced Graphite Creep 3 (AGC-3), began with Advanced Test Reactor (ATR) Cycle 152B on November 27, 2012, and ended with ATR Cycle 155B on April 23, 2014. This report documents qualification of AGC-3 experiment irradiation monitoring data for use by the Very High Temperature Reactor (VHTR) Technology Development Office (TDO) Program for research and development activities required to design and license the first VHTR nuclear plant. Qualified data meet the requirements for data collection and use as described in the experiment planning and quality assurance documents. Failed data do not meet the requirements. Trend data may not meet the requirements, but may still provide some useable information. The report documents qualification of AGC-3 experiment irradiation monitoring data following MCP-2691. This report also documents whether AGC-3 experiment irradiation monitoring data meet the requirements for data collection as specified in technical and functional requirements documents and quality assurance (QA) plans. Data handling is described showing how data are passed from the data collection experiment to the Nuclear Data Management and Analysis System (NDMAS) team. The data structure is described, including data batches, components, attributes, and response variables. The description of the approach to data qualification includes the steps taken to qualify the data and the specific tests used to verify that the data meet requirements. Finally, the current status of the data received by NDMAS from the AGC-3 experiment is presented with summarized information on test results and resolutions. This report addresses all of the irradiation monitoring data collected during the AGC-3 experiment.

  7. Creep behavior for advanced polycrystalline SiC fibers

    International Nuclear Information System (INIS)

    Youngblood, G.E.; Jones, R.H.; Kohyama, Akira

    1997-01-01

    A bend stress relaxation (BSR) test is planned to examine irradiation enhanced creep in polycrystalline SiC fibers which are under development for use as fiber reinforcement in SiC/SiC composite. Baseline 1 hr and 100 hr BSR thermal creep open-quotes mclose quotes curves have been obtained for five selected advanced SiC fiber types and for standard Nicalon CG fiber. The transition temperature, that temperature where the S-shaped m-curve has a value 0.5, is a measure of fiber creep resistance. In order of decreasing thermal creep resistance, with the 100 hr BSR transition temperature given in parenthesis, the fibers ranked: Sylramic (1261 degrees C), Nicalon S (1256 degrees C), annealed Hi Nicalon (1215 degrees C), Hi Nicalon (1078 degrees C), Nicalon CG (1003 degrees C) and Tyranno E (932 degrees C). The thermal creep for Sylramic, Nicalon S, Hi Nicalon and Nicalon CG fibers in a 5000 hr irradiation creep BSR test is projected from the temperature dependence of the m-curves determined during 1 and 100 hr BSR control tests

  8. Creep behavior for advanced polycrystalline SiC fibers

    Energy Technology Data Exchange (ETDEWEB)

    Youngblood, G.E.; Jones, R.H. [Pacific Northwest National Lab., Richland, WA (United States); Kohyama, Akira [Kyoto Univ. (Japan)] [and others

    1997-08-01

    A bend stress relaxation (BSR) test is planned to examine irradiation enhanced creep in polycrystalline SiC fibers which are under development for use as fiber reinforcement in SiC/SiC composite. Baseline 1 hr and 100 hr BSR thermal creep {open_quotes}m{close_quotes} curves have been obtained for five selected advanced SiC fiber types and for standard Nicalon CG fiber. The transition temperature, that temperature where the S-shaped m-curve has a value 0.5, is a measure of fiber creep resistance. In order of decreasing thermal creep resistance, with the 100 hr BSR transition temperature given in parenthesis, the fibers ranked: Sylramic (1261{degrees}C), Nicalon S (1256{degrees}C), annealed Hi Nicalon (1215{degrees}C), Hi Nicalon (1078{degrees}C), Nicalon CG (1003{degrees}C) and Tyranno E (932{degrees}C). The thermal creep for Sylramic, Nicalon S, Hi Nicalon and Nicalon CG fibers in a 5000 hr irradiation creep BSR test is projected from the temperature dependence of the m-curves determined during 1 and 100 hr BSR control tests.

  9. Chemical atomization of graphite by H+ ions

    International Nuclear Information System (INIS)

    Busharov, I.P.; Gorbatov, E.A.; Gusev, V.M.; Guseva, M.I.; Martynenko, Yu.V.

    A simple model of the mechanism of chemical atomization is given, on whose basis a decrease in chemical atomization is qualitatively predicted for high temperatures. Mass spectrometric investigations of the atomization products cited, which found CH 4 and CH 3 molecules during the irradiation of graphite and H + ions thereby confirmed the presence of chemical atomization. A relationship of S and temperature of graphite T during irradiation was obtained which showed a decrease in the coefficient of atomization of a high temperature. (U.S.)

  10. Graphite target for the spiral project

    International Nuclear Information System (INIS)

    Putaux, J.C.; Ducourtieux, M.; Ferro, A.; Foury, P.; Kotfila, L.; Mueller, A.C.; Obert, J.; Pauwels, N.; Potier, J.C.; Proust, J.; Loiselet, M.

    1996-01-01

    A study of the thermal and physical properties of graphite targets for the SPIRAL project is presented. The main objective is to develop an optimized set-up both mechanically and thermally resistant, presenting good release properties (hot targets with thin slices). The results of irradiation tests concerning the mechanical and thermal resistance of the first prototype of SPIRAL target with conical geometry are presented. The micro-structural properties of the graphite target is also studied, in order to check that the release properties are not deteriorated by the irradiation. Finally, the results concerning the latest pilot target internally heated by an electrical current are shown. (author)

  11. Impermeable Graphite: A New Development for Embedding Radioactive Waste

    International Nuclear Information System (INIS)

    Fachinger, Johannes

    2016-01-01

    Irradiated graphite has to be handled as radioactive waste after the operational period of the reactor. However, the waste management of irradiated graphite e.g. from the Spanish Vandellos reactor shows, that waste management of even low contaminated graphite could be expensive and requires special retrieval, treatment and disposal technologies for safe long term storage as low or medium radioactive waste. FNAG has developed an impermeable graphite matrix (IGM) as nuclear waste embedding material. This IGM provides a long term stable enclosure of radioactive waste and can reuse irradiated graphite as feedstock material. Therefore, no additional disposal volume is required if e.g. concrete waste packages were replaced by IGM waste packages. The variability of IGM as embedding has been summarized in the following paper usable for metal scraps, ion exchange resins or debris from buildings. Furthermore the main physical, chemical and structural properties are described. (author)

  12. Computation of deformations and stresses in graphite blocks for HTR core survey purposes

    International Nuclear Information System (INIS)

    Besdo, Dieter; Theymann, W.

    1975-01-01

    Stresses and deformations in graphite fuel elements for HTRs are caused by the temperature distribution and by irradiation under influence of creep, shrinking, thermal strains, and elastic deformations. The global deformations and the stress distribution in a prismatic fuel-element containing regularly distributed axial holes for the coolant flow and the fuel sticks, can be computed in the following manner: the block with its holes is treated as an effective homogeneous continuum with an equivalent global behaviour. Assuming that the fourth-order-tensor of the elastic constants is proportional to the corresponding tensor in the constitutive equations for creep, only the effective strains are of interest. The values of temperature and dose may be given in n points of the block at certain points of time. Then, the inelastic nonthermal strains are integrated by a Runge-Kutta-procedure in the n points. When interpolated and combined with thermal strains, they are incompatible. Hence, they produce elastic deformations which cause creep and can be computed by use of a Ritz-polynomial-series by help of a specific principle of the minimum of potential energy. Excessive computing time can be avoided easily since the influence of the local variation of the elastic constants within the block is almost negligible and, therefore, of practically no importance for the determination of the elastic strains. By this reason some matrices can be calculated a priori, and the elastic deformations are obtained by multiplications of these matrices rather than inversions. Therefore, this method is particularly suited for the computation of deformations and stresses for reactor core survey purposes where a large number (up to 7000 blocks) have to be treated

  13. Creep and creep rupture properties of cladding tube (type 316) in high temperature sodium

    International Nuclear Information System (INIS)

    Atsumo, H.

    1977-01-01

    The thin walled small sized seamless AISI 316 steel tubes, which are designated to be domestically used as the fuel cladding tube for sodium cooled fast breeder reactors in Japan, are irradiated in the following sodium of high temperature in the range of 370 deg. C to 700 deg. C, and receive gradually increased internal pressure caused by the fission produced gas generating from the nuclear fuel burn-up inside the cladding tube. Consequently, the creep behavior of fuel cladding tubes under a high temperature sodium environment is an important problem which must be determined and clarified together with their characteristic features under irradiation and in air. In relation to the creep performance of fuel cladding tubes made of AISI 316 steel and other comparable austenitic stainless steels, hardly any studies are found that are made systematically to examine the effect of sodium with sodium purity as parameter or any comparative studies with in-air data at various different temperatures. The present research work was aimed to obtain certain basic design data relating to in-sodium creep performance of the domestic made fuel cladding tubes for fast breeder reactors, and also to gain further date as considered necessary under several sodium conditions. That is, together with establishment of the technology for tensile creep test and internal pressure creep rupture test in flowing sodium of high temperature, a series of tests and studies were performed on the trial made cladding tubes of AISI Type-316 steel. In the first place, two kinds of purity conditions of sodium, close to the actual reactor-operating condition, (oxygen concentration of 10 ppm and 5 ppm respectively) were established, and then uniaxial tensile creep test and rupture test under various temperatures were performed and the resulting data were compared and evaluated against the in-air data. Then, secondly, an internal pressure creep rupture test was conducted under a single purity sodium environment

  14. Micro creep mechanisms of tungsten

    International Nuclear Information System (INIS)

    Levoy, R.; Hugon, I.; Burlet, H.; Baillin, X.; Guetaz, L.

    2000-01-01

    Due to its high melting point (3410 deg C), tungsten offers good mechanical properties at elevated temperatures for several applications in non-oxidizing environment. The creep behavior of tungsten is well known between 1200 and 2500 deg C and 10 -3 to 10 -1 strain. However, in some applications when dimensional stability of components is required, these strains are excessive and it is necessary to know the creep behavior of the material for micro-strains (between 10 -4 and 10 -6 ). Methods and devices used to measure creep micro-strains are presented, and creep equations (Norton and Chaboche laws) were developed for wrought, annealed and recrystallized tungsten. The main results obtained on tungsten under low stresses are: stress exponent 1, symmetry of micro-strains in creep-tension and creep-compression, inverse creep (threshold stress), etc. TEM, SEM and EBSD studies allow interpretation of the micro-creep mechanism of tungsten under low stresses and low temperature (∼0.3 K) like the Harper-Dorn creep. In Harper-Dorn creep, micro-strains are associated with the density and the distribution of dislocations existing in the crystals before creep. At 975 deg C, the initial dislocation structure moves differently whether or not a stress is applied. To improve the micro-creep behavior of tungsten, a heat treatment is proposed to create the optimum dislocation structure. (authors)

  15. AGC-4 Experiment Irradiation Monitoring Data Qualification Interim Report

    International Nuclear Information System (INIS)

    Hull, Laurence Charles

    2016-01-01

    The Graphite Technology Development Program is running a series of six experiments to quantify the effects of irradiation on nuclear grade graphite. The fourth experiment, Advanced Graphite Creep 4 (AGC 4), began with Advanced Test Reactor (ATR) cycle 157D on May 30, 2015, and has been irradiated for two cycles. The capsule was removed from the reactor after ATR cycle 158A, which ended on January 2, 2016, due to interference with another experiment. Irradiation will resume when the interfering experiment is removed from the reactor. This report documents qualification of AGC 4 experiment irradiation monitoring data for use by the Advanced Reactor Technologies (ART) Technology Development Office (TDO) Program for research and development activities required to design and license the first HTR nuclear plant. Qualified data meet the requirements for use as described in the experiment planning and quality assurance documents. Failed data do not meet the requirements and provide no useable information. Trend data may not meet all requirements, but still provide some useable information. Use of Trend data requires assessment of how any deficiencies affect a particular use of the data. All thermocouples (TCs) have functioned throughout the AGC-4 experiment. All temperature data are Qualified for use by the ART TDO Program. Argon, helium, and total gas flow data were within expected ranges and are Qualified for use by the ART TDO Program. Discharge gas line moisture values were consistently low during cycle 157D. At the start of cycle 158A, gas moisture briefly spiked to over 600 ppmv and then declined throughout the cycle. Moisture values are within the measurement range of the instrument and are Qualified for use by the ART TDO Program. Graphite creep specimens were subjected to one of three loads, 393, 491, or 589 lbf. For a brief period during cycle 157D between 12:19 on June 2, 2015 and 08:23 on June 11, 2015 the load cells were wired incorrectly resulting in missing

  16. AGC-4 Experiment Irradiation Monitoring Data Qualification Interim Report

    Energy Technology Data Exchange (ETDEWEB)

    Hull, Laurence Charles [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-08-01

    The Graphite Technology Development Program is running a series of six experiments to quantify the effects of irradiation on nuclear grade graphite. The fourth experiment, Advanced Graphite Creep 4 (AGC 4), began with Advanced Test Reactor (ATR) cycle 157D on May 30, 2015, and has been irradiated for two cycles. The capsule was removed from the reactor after ATR cycle 158A, which ended on January 2, 2016, due to interference with another experiment. Irradiation will resume when the interfering experiment is removed from the reactor. This report documents qualification of AGC 4 experiment irradiation monitoring data for use by the Advanced Reactor Technologies (ART) Technology Development Office (TDO) Program for research and development activities required to design and license the first HTR nuclear plant. Qualified data meet the requirements for use as described in the experiment planning and quality assurance documents. Failed data do not meet the requirements and provide no useable information. Trend data may not meet all requirements, but still provide some useable information. Use of Trend data requires assessment of how any deficiencies affect a particular use of the data. All thermocouples (TCs) have functioned throughout the AGC-4 experiment. All temperature data are Qualified for use by the ART TDO Program. Argon, helium, and total gas flow data were within expected ranges and are Qualified for use by the ART TDO Program. Discharge gas line moisture values were consistently low during cycle 157D. At the start of cycle 158A, gas moisture briefly spiked to over 600 ppmv and then declined throughout the cycle. Moisture values are within the measurement range of the instrument and are Qualified for use by the ART TDO Program. Graphite creep specimens were subjected to one of three loads, 393, 491, or 589 lbf. For a brief period during cycle 157D between 12:19 on June 2, 2015 and 08:23 on June 11, 2015 the load cells were wired incorrectly resulting in missing

  17. Effect of graphite reflector on activation of fusion breeding blanket

    International Nuclear Information System (INIS)

    Lee, Cheol Woo; Lee, Young-Ouk; Lee, Dong Won; Cho, Seungyon; Ahn, Mu-Young

    2016-01-01

    Highlights: • The graphite reflector concept has been applied in the design of the Korea HCCR TBM for ITER and this concept is also a candidate design option for Korea Demo. • In the graphite reflector, C-14, B-11 and Be-10 are produced after an irradiation. Impurities in both case of beryllium and graphite is dominant in the shutdown dose after an irradiation. • Based on the evaluation, the graphite reflector is a good alternative of the beryllium multiplier in the view of induced activity and shutdown dose. But C-14 produced in the graphite reflector should be considered carefully in the view of radwaste management. - Abstract: Korea has proposed a Helium-Cooled Ceramic Reflector (HCCR) breeding blanket concept relevant to fusion power plants. Here, graphite is used as a reflector material by reducing the amount of beryllium multiplier. In this paper, activity analysis was performed and the effect of graphite reflector in the view of activation was compared to the beryllium multiplier. As a result, it is expected that using the graphite reflector instead of the beryllium multiplier decreases total activity very effectively. But the graphite reflector produces C-14 about 17.2 times than the beryllium multiplier. Therefore, C-14 produced in the graphite reflector is expected as a significant nuclide in the view of radwaste management.

  18. In-pile Creep Tests of Zircaloy Tubing in the Studsvik R2 Reactor. Final Report

    International Nuclear Information System (INIS)

    Tomani, Hans; Lindeloew, Ulf

    2000-12-01

    In this report are presented the findings of a prototype creep test on Zr4 guide tube specimens exposed in-pile and out-of-pile and stressed by constant bending moments. The calculated initial deflection curvature caused by the applied bending moment agrees very well with the measured initial values. Furthermore, the measurement results show excellent consistency. The dominant impact of neutron irradiation is clearly demonstrated. After 3 cycles (∼1300 hours) the irradiation creep is 4 times as large as the thermal creep. This is the case at least when fresh tube material is used. Irradiation creep progresses steadily, but the creep rate is not quite constant during the 3 irradiation cycles. The thermal creep, on the other hand, quickly saturates and there is hardly any further deflection after the second cycle for the specimen situated above the core. A limitation with the rig has been that the tube deflection became limited by the rig carrier body of the rig in the neutron flux (core) that disqualified the results of a fourth irradiation cycle actually performed in the fall of 1998. The test method appears to be suitable for testing the bending creep of different guide tube materials or designs under PWR conditions

  19. Magnetic order in graphite: Experimental evidence, intrinsic and extrinsic difficulties

    International Nuclear Information System (INIS)

    Esquinazi, P.; Barzola-Quiquia, J.; Spemann, D.; Rothermel, M.; Ohldag, H.; Garcia, N.; Setzer, A.; Butz, T.

    2010-01-01

    We discuss recently obtained data using different experimental methods including magnetoresistance measurements that indicate the existence of metal-free high-temperature magnetic order in graphite. Intrinsic as well as extrinsic difficulties to trigger magnetic order by irradiation of graphite are discussed in view of recently published theoretical work.

  20. High-temperature transient creep properties of CANDU pressure tubes

    International Nuclear Information System (INIS)

    Fong, R.W.L.; Chow, C.K.

    2002-06-01

    During a hypothetical large break loss-of-coolant accident (LOCA), the coolant flow would be reduced in some fuel channels and would stagnate and cause the fuel temperature to rise and overheat the pressure tube. The overheated pressure tube could balloon (creep radially) into contact with its moderator-cooled calandria tube. Upon contact, the stored thermal energy in the pressure tube is transferred to the calandria tube and into the moderator, which acts as a heat sink. For safety analyses, the modelling of fuel channel deformation behaviour during a large LOCA requires a sound knowledge of the high-temperature creep properties of Zr-2.5Nb pressure tubes. To this extent, a ballooning model to predict pressure-tube deformation was developed by Shewfelt et al., based on creep equations derived using uniaxial tensile specimens. It has been recognized, however, that there is an inherent variability in the high-temperature creep properties of CANDU pressure tubes. The variability, can be due to different tube-manufacturing practices, variations in chemical compositions, and changes in microstructure induced by irradiation during service in the reactor. It is important to quantify the variability of high-temperature creep properties so that accurate predictions on pressure-tube creep behaviour can be made. This paper summarizes recent data obtained from high-temperature uniaxial creep tests performed on specimens taken from both unirradiated (offcut) and irradiated pressure tubes, suggesting that the variability is attributed mainly to the initial differences in microstructure (grain size, shape and preferred orientation) and also from tube-to-tube variations in chemical composition, rather than due to irradiation exposure. These data will provide safety analysts with the means to quantify the uncertainties in the prediction of pressure-tube contact temperatures during a postulated large break LOCA. (author)

  1. Consistent creep and rupture properties for creep-fatigue evaluation

    International Nuclear Information System (INIS)

    Schultz, C.C.

    1978-01-01

    The currently accepted practice of using inconsistent representations of creep and rupture behaviors in the prediction of creep-fatigue life is shown to introduce a factor of safety beyond that specified in current ASME Code design rules for 304 stainless steel Class 1 nuclear components. Accurate predictions of creep-fatigue life for uniaxial tests on a given heat of material are obtained by using creep and rupture properties for that same heat of material. The use of a consistent representation of creep and rupture properties for a mininum strength heat is also shown to provide adequate predictions. The viability of using consistent properties (either actual or those of a minimum heat) to predict creep-fatigue life thus identifies significant design uses for the results of characterization tests and improved creep and rupture correlations

  2. Consistent creep and rupture properties for creep-fatigue evaluation

    International Nuclear Information System (INIS)

    Schultz, C.C.

    1979-01-01

    The currently accepted practice of using inconsistent representations of creep and rupture behaviors in the prediction of creep-fatigue life is shown to introduce a factor of safety beyond that specified in current ASME Code design rules for 304 stainless steel Class 1 nuclear components. Accurate predictions of creep-fatigue life for uniaxial tests on a given heat of material are obtained by using creep and rupture properties for that same heat of material. The use of a consistent representation of creep and rupture properties for a minimum strength heat is also shown to provide reasonable predictions. The viability of using consistent properties (either actual or those of a minimum strength heat) to predict creep-fatigue life thus identifies significant design uses for the results of characterization tests and improved creep and rupture correlations. 12 refs

  3. Mechanisms of radiation induced creep and growth

    International Nuclear Information System (INIS)

    Bullough, R.; Wood, M.H.

    1980-01-01

    Irradiation creep occurs primarily because the applied stress causes the evolving microstructure to respond in an anisotropic fashion to the interstitial and vacancy fluxes. On the other hand, irradiation growth requires the response to be naturally anisotropic in the absence of applied stress. Four fundamental mechanisms of irradiation creep have been conjectured: stress induced preferred absorption (SIPA) of the point defects on the dislocations, stress induced preferred nucleation (SIPN) of point defects in planar aggregates (edge dislocation loops), stress induced climb and glide (SICG) of the dislocation network and stress induced gas driven interstitial deposition (SIGD). These mechanisms will be briefly outlined and commented upon. The contributions made by these mechanisms to the total strain are not, in general, mutually separable and also depend on the prevailing (and changing) microstructure during irradiation. The fundamental mechanism of irradiation growth will be discussed: it is believed to arise by the preferred condensation of point defects and climb of dislocation loops and network on certain crystallographic planes. The preferred absorption and nucleation is thus a consequence of natural crystallographic anisotropy and not due to any external stresses. Again the effectiveness of this mechanism depends on the prevailing microstructure in the material. In this connection will be particularly drawn to the significance of solute trapping, segregation at grain boundaries, dislocation bias for interstitials and transport parameters for an understanding of irradiation growth in materials like zirconium and its alloys; the relevance of recent simulation studies of growth in such materials using electrons to the growth under neutron irradiation will be discussed in detail and a consistent model of growth in these materials will be presented. (orig.)

  4. Theoretical basis for graphite stress analysis in BERSAFE

    International Nuclear Information System (INIS)

    Harper, P.G.

    1980-03-01

    The BERSAFE finite element computer program for structural analysis has been extended to deal with structures made from irradiated graphite. This report describes the material behaviour which has been modelled and gives the theoretical basis for the solution procedure. (author)

  5. Analysis of picosecond pulsed laser melted graphite

    International Nuclear Information System (INIS)

    Steinbeck, J.; Braunstein, G.; Speck, J.; Dresselhaus, M.S.; Huang, C.Y.; Malvezzi, A.M.; Bloembergen, N.

    1986-01-01

    A Raman microprobe and high resolution TEM have been used to analyze the resolidified region of liquid carbon generated by picosecond pulse laser radiation. From the relative intensities of the zone center Raman-allowed mode for graphite at 1582 cm -1 and the disorder-induced mode at 1360 cm -1 , the average graphite crystallite size in the resolidified region is determined as a function of position. By comparison with Rutherford backscattering spectra and Raman spectra from nonosecond pulsed laser melting experiments, the disorder depth for picosecond pulsed laser melted graphite is determined as a function of irradiating energy density. Comparisons of TEM micrographs for nanosecond and picosecond pulsed laser melting experiments show that the structure of the laser disordered regions in graphite are similar and exhibit similar behavior with increasing laser pulse fluence

  6. Creep of timber joints

    NARCIS (Netherlands)

    Van de Kuilen, J.W.G.

    2008-01-01

    A creep analysis has been performed on nailed, toothed-plates and split-ring joints in a varying uncontrolled climate. The load levels varied between 30% and 50% of the average ultimate short term strength of these joints, tested in accordance with ISO 6891. The climate in which the tests were

  7. Analysis of indentation creep

    Science.gov (United States)

    Don S. Stone; Joseph E. Jakes; Jonathan Puthoff; Abdelmageed A. Elmustafa

    2010-01-01

    Finite element analysis is used to simulate cone indentation creep in materials across a wide range of hardness, strain rate sensitivity, and work-hardening exponent. Modeling reveals that the commonly held assumption of the hardness strain rate sensitivity (mΗ) equaling the flow stress strain rate sensitivity (mσ...

  8. In-pile creep test technique for zirconium alloys examination in BR-10 reactor channels

    International Nuclear Information System (INIS)

    Pevchikh, Yu.M.; Kruglov, A.S.; Troyanov, V.M.

    2002-01-01

    The irradiation enhanced creep phenomenon was discovered in stainless steels as a specific physical process accompanying high-intensity neutron flux irradiation in fast reactors. IPPE is also experienced in irradiation creep test activities, studying different types of materials under irradiation in BR-10 fast reactor. Series of in-channel type test facilities were constructed and tested in BR-10 reactor's 'dry' channels in order to carry out full-scale instrumented examination regarded to in-pile creep behaviour of different reactor materials. As a result, a specific test technique, named 'Tensometric method', has been developed and experimentally proved to be power enough in order to investigate irradiation creep of materials right in situ under neutron irradiation. The main peculiarity of test facility, which is constructed to apply the tensometric method, consists in absence of any special deformation-measurement cell at all. The in-pile creep strain measurement technique developed at IPPE is based on the non-direct measurement of specimen's deformation (either linear tensile strain or angular twisting one), which directly affects the loaded draws' tension parameters. Starting from 1993, in-pile creep experiments to investigate in-reactor creep behaviour of E110 and E635 zirconium alloys were carried out in BR-10. Experimental results and data collected during more than 20-year of BR-10 in-reactor creep test experience can be assumed as a strong evidence that the tensometric technique is a powerful instrument, which can give a chance to study different irradiation effects on reactor materials directly under irradiation. (author)

  9. Creep buckling of shell structures

    International Nuclear Information System (INIS)

    Miyazaki, Noriyuki; Hagihara, Seiya

    2015-01-01

    The present article contains a review of the literatures on the creep buckling of shell structures published from late 1950's to recent years. In this article, the creep buckling studies on circular cylindrical shells, spherical shells, partial cylindrical shells and other shells are reviewed in addition to creep buckling criteria. Creep buckling is categorized into two types. One is the creep buckling due to quasi-static instability, in which the critical time for creep buckling is determined by tracing a creep deformation versus time curve. The other is the creep buckling due to kinetic instability, in which the critical time can be determined by examining the shape of total potential energy in the vicinity of a quasi-static equilibrium state. Bifurcation buckling and snap-through buckling during creep deformation belong to this type of creep buckling. A few detailed descriptions are given to the bifurcation and snap-through type of creep buckling based on the present authors' works. (author)

  10. Strength and life under creeping

    International Nuclear Information System (INIS)

    Pospishil, B.

    1982-01-01

    Certain examples of the application of the Lepin modified creep model, which are of interest from technical viewpoint, are presented. Mathematical solution of the dependence of strength limit at elevated temperatures on creep characteristics is obtained. Tensile test at elevated temperatures is a particular case of creep or relaxation and both strength limit and conventional yield strength at elevated temperatures are completely determined by parameters of state equations during creep. The equation of fracture summing during creep is confirmed not only by the experiment data when stresses change sporadically, but also by good reflection of durability curve using the system of equations. The system presented on the basis of parameters of the equations obtained on any part of durability curve, permits to forecast the following parameters of creep: strain, strain rate, life time, strain in the process of fracture. Tensile test at elevated temperature is advisable as an addition when determining creep curves (time-strain curves) [ru

  11. Creep behaviour and creep mechanisms of normal and healing ligaments

    Science.gov (United States)

    Thornton, Gail Marilyn

    Patients with knee ligament injuries often undergo ligament reconstructions to restore joint stability and, potentially, abate osteoarthritis. Careful literature review suggests that in 10% to 40% of these patients the graft tissue "stretches out". Some graft elongation is likely due to creep (increased elongation of tissue under repeated or sustained load). Quantifying creep behaviour and identifying creep mechanisms in both normal and healing ligaments is important for finding clinically relevant means to prevent creep. Ligament creep was accurately predicted using a novel yet simple structural model that incorporated both collagen fibre recruitment and fibre creep. Using the inverse stress relaxation function to model fibre creep in conjunction with fibre recruitment produced a superior prediction of ligament creep than that obtained from the inverse stress relaxation function alone. This implied mechanistic role of fibre recruitment during creep was supported using a new approach to quantify crimp patterns at stresses in the toe region (increasing stiffness) and linear region (constant stiffness) of the stress-strain curve. Ligament creep was relatively insensitive to increases in stress in the toe region; however, creep strain increased significantly when tested at the linear region stress. Concomitantly, fibre recruitment was evident at the toe region stresses; however, recruitment was limited at the linear region stress. Elevating the water content of normal ligament using phosphate buffered saline increased the creep response. Therefore, both water content and fibre recruitment are important mechanistic factors involved in creep of normal ligaments. Ligament scars had inferior creep behaviour compared to normal ligaments even after 14 weeks. In addition to inferior collagen properties affecting fibre recruitment and increased water content, increased glycosaminoglycan content and flaws in scar tissue were implicated as potential mechanisms of scar creep

  12. Increase of the density of commercial graphite

    International Nuclear Information System (INIS)

    Tobias, H.; Meyerstein, D.

    1977-12-01

    The increase of the density of commercial graphite of the type ATJ by polymerization of an impregnated monomer, followed by pyrolysis, is described. The monomer which was either styrene or acrylonitrile, was irradiated by a 60 Co source and pyrolized in a standard vacuum system. The irradiation dose for the polymerization of the monomer was determined. Suggestions for the establishment of the optimum conditions are offered

  13. Effect of gamma radiation on graphite - PTFE dry lubrication system

    Science.gov (United States)

    Singh, Sachin; Tyagi, Mukti; Seshadri, Geetha; Tyagi, Ajay Kumar; Varshney, Lalit

    2017-12-01

    An effect of gamma radiation on lubrication behavior of graphite -PTFE dry lubrication system has been studied using (TR-TW-30L) tribometer with thrust washer attachment in plane contact. Different compositions of graphite and PTFE were prepared and irradiated by gamma rays. Gamma radiation exposure significantly improves the tribological properties indicated by decrease in coefficient of friction and wear properties of graphite -PTFE dry lubrication system. SEM and XRD analysis confirm the physico-chemical modification of graphite-PTFE on gamma radiation exposure leading to a novel dry lubrication system with good slip and anti friction properties.

  14. Anisotropic thermal creep of internally pressurized Zr-2.5Nb tubes

    International Nuclear Information System (INIS)

    Li, W.; Holt, R.A.

    2010-01-01

    The anisotropy of creep of internally pressurized cold-worked Zr-2.5Nb tubes with different crystallographic textures is reported. The stress exponent n was determined to be about three at transverse stresses from 100 to 250 MPa with an activation energy of ∼99.54 kJ/mol in the temperature range 300-400 o C. The stress exponent increased to ∼6 for transverse stresses from 250 to 325 MPa. From this data an experimental regime of 350 o C and 300 MPa was established in which dislocation glide is the likely strain-producing mechanism. Creep tests were carried out under these conditions on internally pressurized Zr-2.5Nb tubes with 18 different textures. Creep strain and creep anisotropy (ratio of axial to transverse steady-state creep rate, ε . A /ε . T ) exhibited strong dependence on crystallographic textures of the Zr-2.5Nb tubes. It was found that the values of (ε . A /ε . T ) increased as the difference between the resolved faction of basal plane normals in the transverse and radial directions (f T - f R ) increases. The tubes with the strongest radial texture showed a negative axial creep strain and a negative creep rate ratio (ε . A /ε . T ) and tubes with a strong transverse texture exhibited the positive values of steady-state creep rate ratio (ε . A /ε . T ) and good creep resistance in the transverse direction. These behaviors are qualitatively similar to those observed during irradiation creep, and also to the predictions of polycrystalline models for creep in which glide is the strain-producing mechanism and prismatic slip is the dominant system. A detailed analysis of the results using polycrystalline models may assist in understanding the anisotropy of irradiation creep.

  15. Development of design data for graphite reinforced epoxy and polyimide composites

    Science.gov (United States)

    Scheck, W. G.

    1974-01-01

    Processing techniques and design data were characterized for a graphite/epoxy composite system that is useful from 75 K to 450 K, and a graphite/polyimide composite system that is useful from 75 K to 589 K. The Monsanto 710 polyimide resin was selected as the resin to be characterized and used with the graphite fiber reinforcement. Material was purchased using the prepreg specification for the design data generation for both the HT-S/710 and HM-S/710 graphite/polyimide composite system. Lamina and laminate properties were determined at 75 K, 297 K, and 589 K. The test results obtained on the skin-stringer components proved that graphite/polyimide composites can be reliably designed and analyzed much like graphite/epoxy composites. The design data generated in the program includes the standard static mechanical properties, biaxial strain data, creep, fatigue, aging, and thick laminate data.

  16. Microstructure in Zircaloy Creep Tested in the R2 Reactor

    International Nuclear Information System (INIS)

    Pettersson, Kjell

    2004-12-01

    Tubular specimens of Zircaloy-4 have been creep tested in bending in the R2 reactor in Studsvik. The creep deformation in the reactor core is accelerated in comparison with creep deformation outside the reactor core. The possible mechanisms behind this behaviour are described briefly. In order to determine which the actual mechanism is, the microstructure of the material creep tested in the R2 reactor has been examined by transmission electron microscopy. Due to the bending, material subjected to both tensile and compressive stress during creep was available. Since some of the proposed mechanisms might give microstructures which are different when the material is subjected to compressive or tensile stress it was assumed that examination of both types of material would give valuable information with regard to the operating mechanism. The result of the examination was that in the as-irradiated condition there were no obvious differences detected between materials which had been deformed in tension or compression. After a heat treatment to coarsen the irradiation induced microstructure there were still no significant differences between the two types of material. However it was now observed that in addition to dislocation loops the microstructure also contained network dislocations which presumably had been invisible in the electron microscope before heat treatment due to the high density of small dislocation loops in this state. It is therefore concluded that the most probable mechanism for irradiation creep in this case is climb and glide of the network dislocations. The role of irradiation is two-fold: It accelerates climb due to the production of point defects of which more interstitials than vacancies arrive to the network dislocations stopped at an obstacles. This leads to a net climb after which a dislocation is released from the obstacle and an amount of glide takes place. The second effect is the production of loops which serve as an increasing density of

  17. Characteristics of first loaded IG-110 graphite in HTTR core

    International Nuclear Information System (INIS)

    Sumita, Junya; Shibata, Taiju; Iyoku, Tatsuo; Sawa, Kazuhiro; Hanawa, Satoshi; Ishihara, Masahiro

    2006-10-01

    IG-110 graphite is a fine-grained isotropic and nuclear-grade graphite with excellent resistivity on both irradiation and corrosion and with high strength. The IG-110 graphite is used for the graphite components of High Temperature Engineering Test Reactor (HTTR) such as fuel and control rod guide blocks and support posts. In order to design and fabricate the graphite components in the HTTR, the Japan Atomic Energy Research Institute (the Japan Atomic Energy Agency at present) had established the graphite structural design code and design data on the basis of former research results. After the design code establishment, the IG-110 graphite components were fabricated and loaded in the HTTR core. This report summarized the characteristics of the first loaded IG-110 graphite as basic data for surveillance test, measuring material characteristics changed by neutron irradiation and oxidation. By comparing the design data, it was shown that the first loaded IG-110 graphite had excellent strength properties and enough safety margins to the stress limits in the design code. (author)

  18. Graphite moderator lifecycle behaviour. Proceedings of a specialists meeting

    International Nuclear Information System (INIS)

    1996-08-01

    The meeting provided the forum for graphite specialists representing operating and research organizations worldwide to exchange information in the following areas: the status of graphite development; operation and safety procedures for existing and future graphite moderated reactors; graphite testing techniques; review of the experiences gained and data acquired on the influence of neutron irradiation and oxidizing conditions on key graphite properties; and to exchange information useful for decommissioning activities. The participants provided twenty-seven papers on behalf of their countries and respective technical organizations. An open discussion followed each of the presentations. A consistently reoccurring theme throughout the specialists meeting was the noticeable reduction in the number of graphite experts remaining the nuclear power industry. Graphite moderated power reactors have provided a significant contribution to the generation of electricity throughout the past forty years and will continue to be a prominent energy source for the future. Yet, many of the renowned experts in the field of nuclear graphites are nearing the end of their careers without apparent replacement. This, coupled with changes in the focus on nuclear power by some industrialized countries, has prompted the IAEA to initiate an evaluation on the feasibility and interest by Member States of establishing a central archive facility for the storage of data on irradiated graphites. Refs, figs, tabs

  19. Graphite moderator lifecycle behaviour. Proceedings of a specialists meeting

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-08-01

    The meeting provided the forum for graphite specialists representing operating and research organizations worldwide to exchange information in the following areas: the status of graphite development; operation and safety procedures for existing and future graphite moderated reactors; graphite testing techniques; review of the experiences gained and data acquired on the influence of neutron irradiation and oxidizing conditions on key graphite properties; and to exchange information useful for decommissioning activities. The participants provided twenty-seven papers on behalf of their countries and respective technical organizations. An open discussion followed each of the presentations. A consistently reoccurring theme throughout the specialists meeting was the noticeable reduction in the number of graphite experts remaining the nuclear power industry. Graphite moderated power reactors have provided a significant contribution to the generation of electricity throughout the past forty years and will continue to be a prominent energy source for the future. Yet, many of the renowned experts in the field of nuclear graphites are nearing the end of their careers without apparent replacement. This, coupled with changes in the focus on nuclear power by some industrialized countries, has prompted the IAEA to initiate an evaluation on the feasibility and interest by Member States of establishing a central archive facility for the storage of data on irradiated graphites. Refs, figs, tabs.

  20. Creep and creep-rupture behavior of Alloy 718

    International Nuclear Information System (INIS)

    Brinkman, C.R.; Booker, M.K.; Ding, J.L.

    1991-01-01

    Data obtained from creep and creep-rupture tests conducted on 18 heats of Alloy 718 were used to formulate models for predicting high temperature time dependent behavior of this alloy. Creep tests were conducted on specimens taken from a number of commercial product forms including plate, bar, and forgoing material that had been procured and heat treated in accordance with ASTM specifications B-670 or B-637. Data were obtained over the temperature range of 427 to 760 degree C ad at test times to about 87,000 h. Comparisons are given between experimental data and the analytical models. The analytical models for creep-rupture included one based on lot-centering regression analysis and two based on the Minimum Commitment Method. A ''master'' curve approach was used to develop and equation for estimating creep deformation up to the onset of tertiary creep. 11 refs., 13 figs

  1. Description of Concrete Creep under Time-Varying Stress Using Parallel Creep Curve

    OpenAIRE

    Park, Yeong-Seong; Lee, Yong-Hak; Lee, Youngwhan

    2016-01-01

    An incremental format of creep model was presented to take account of the development of concrete creep due to loading at different ages. The formulation was attained by introducing a horizontal parallel assumption of creep curves and combining it with the vertical parallel creep curve of the rate of creep method to remedy the disadvantage of the rate of creep method that significantly underestimates the amount of creep strain, regardless of its simple format. Two creep curves were combined b...

  2. Physics experiments in graphite lattices (1962); Experiences sur les reseaux a graphite (1962)

    Energy Technology Data Exchange (ETDEWEB)

    Bacher, P; Cogne, F [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1962-07-01

    A review is made of the various experimental methods used to determine the physics of graphite, natural uranium lattices: integral lattice experiments; both absolute and differential, effective cross section measurements, both by activation methods and by analysis of irradiated fuels, fine structure measurements. A number of experimental results are also given. (authors) [French] On decrit les differentes methodes experimentales utilisees pour determiner les parametres physiques de reseaux a uranium-graphite. Il s'agit d'experiences globales: mesures absolues et relatives de laplaciens, mesures de sections efficaces effectives par activation et par analyses de combustibles irradies, mesures de structures fines. Un certain nombre de resultats experimentaux sont communiques. (auteurs)

  3. Remaining life assessment of carbon steel boiler headers by repeated creep testing

    Energy Technology Data Exchange (ETDEWEB)

    Drew, M. [ANSTO, Materials and Engineering Science, New Illawarra Road, Lucas Heights, PMB 1 Menai, NSW 2234 (Australia)]. E-mail: michael.drew@ansto.gov.au; Humphries, S. [ANSTO, Materials and Engineering Science, New Illawarra Road, Lucas Heights, PMB 1 Menai, NSW 2234 (Australia); Thorogood, K. [ANSTO, Materials and Engineering Science, New Illawarra Road, Lucas Heights, PMB 1 Menai, NSW 2234 (Australia); Barnett, N. [BlueScope Steel, P.O. Box 1854, Wollongong, NSW (Australia)

    2006-05-15

    The condition of carbon steel boiler headers that have been in service for over 25 years has been assessed periodically by NDT, dimensional measurements, replication and accelerated creep testing. Historical temperature records were limited, so estimates of effective header temperatures were made from replicas. These estimates were compared with header stub thermocouple readings. At about 280,000 service hours, samples were chain-drilled from the headers for accelerated creep testing. These test results indicated that the headers had satisfactory remaining life. Nine years after the original samples were taken, additional samples were removed from one header at 337,000 service hours. The creep rupture properties measured from the repeated tests were almost identical to the initial results. A mild degree of random, nodular graphite was found in the samples and its effect on creep properties is discussed.

  4. On the separation of so-called non-volatile uranium fission products of uranium using the conversion of neutron-irradiated uranium dioxide and graphite

    International Nuclear Information System (INIS)

    Elhardt, W.

    1979-01-01

    The investigations are continued in the following work which arose from the concept of separating uranium fission products from uranium. This is achieved in that due to the lattice conversions occurring during the course of solid chemical reactions, fission products can easily pass from the uranium-contained solid to a second solid. The investigations carried out primarily concern the release behaviour of cerium and neodymium in the temperature region of 1200 to 1700 0 C. UO 2 + graphite, both in powder form, are selected as suitable reaction system having the preconditions needed for the lattice conversion for the release effect. The target aimed at from the practical aspect for the improved release of lanthanoids is achieved by an isobar test course - changing temperature from 1200 to 1500 0 C at constant pressure, with a cerium release of 75-80% and a neodynium release of 80-90% (maximum at 1400 0 C). The concepts on the mechanism of the fission product release are related to transport processes in crystal lattices, as well as chemical solid reactions and evaporation processes on the surface of UC 2 grains. (orig./RB) [de

  5. Report On Design And Preliminary Data Of Halden In-Pile Creep Rig

    Energy Technology Data Exchange (ETDEWEB)

    Terrani, Kurt A [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Karlsen, T. M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yamamoto, Yukinori [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-01

    A set of in-pile creep tests is ongoing in the Halden reactor on ORNL’s candidate accident tolerant fuel cladding materials. These tests are meant to provide essential material property information that is needed for an informed analysis of these fuel concepts under normal operating conditions. These tests provide detailed information regarding swelling, thermal creep, and irradiation creep rates of these materials. The results to date have been compared with the limited set of information available in literature that is form irradiation tests in other reactors or out-of-pile tests. Most of the results are in good agreement with prior literature, except for irradiation creep rate of SiC. To elucidate the difference between the HFIR and Halden test results continued testing is necessary. The tests describe in this progress report are ongoing and will continue for at least another year.

  6. Creep in sodium

    International Nuclear Information System (INIS)

    Charnock, W.; Cordwell, J.E.

    1978-03-01

    Available information on the creep of austenitic, ferritic and Alloy-800 type steels in liquid sodium is critically reviewed. Creep properties of stainless steels can be affected by element transfer and corrosion. At reactor structural component temperatures environmental effects are likely to be less important than changes due to thermal ageing. At high clad temperatures (700 0 C) decarburisation may cause the loss of strength and ductility in unstabilised steels while cavity formation may cause embrittlement in stabilised steels. The properties of Alloy 800 are, in some experiments, found to deteriorate while in others they are enhanced. This may be a consequence of the metallurgical complexity of the material or arise from the nature of the various techniques employed. Low alloy ferritic steels tend to decarburise in sodium at temperatures greater than 500 0 C and this leads to loss of strength and an increase in ductility. High alloy ferritics are immune to this effect and appear to be able to tolerate a degree of carburisation. Although intergranular cracking may be enhanced in liquid sodium the mechanical consequences are not significant and evidence for the existence of an embrittlement effect not associated with element transfer or corrosion is weak. Stress and strain may enhance element transfer at crack tips. However in real cracks the gettering or supply action of the crack faces conditions the chemistry of the cracks in sodium and protects the crack tip from element transfer. Thus creep crack extension rates should be independent of changes in bulk coolant chemistry. (author)

  7. High temperature creep of vanadium

    International Nuclear Information System (INIS)

    Juhasz, A.; Kovacs, I.

    1978-01-01

    The creep behaviour of polycrystalline vanadium of 99.7% purity has been investigated in the temperature range 790-880 0 C in a high temperature microscope. It was found that the creep properties depend strongly on the history of the sample. To take this fact into account some additional properties such as the dependence of the yield stress and the microhardness on the pre-annealing treatment have also been studied. Samples used in creep measurements were selected on the basis of their microhardness. The activation energy of creep depends on the microhardness and on the creep temperature. In samples annealed at 1250 0 C for one hour (HV=160 kgf mm -2 ) the rate of creep is controlled by vacancy diffusion in the temperature range 820-880 0 C with an activation energy of 78+-8 kcal mol -1 . (Auth.)

  8. Creep of fibrous composite materials

    DEFF Research Database (Denmark)

    Lilholt, Hans

    1985-01-01

    Models are presented for the creep behaviour of fibrous composite materials with aligned fibres. The models comprise both cases where the fibres remain rigid in a creeping matrix and cases where the fibres are creeping in a creeping matrix. The treatment allows for several contributions...... to the creep strength of composites. The advantage of combined analyses of several data sets is emphasized and illustrated for some experimental data. The analyses show that it is possible to derive creep equations for the (in situ) properties of the fibres. The experiments treated include model systems...... such as Ni + W-fibres, high temperature materials such as Ni + Ni3Al + Cr3C2-fibres, and medium temperature materials such as Al + SiC-fibres. For the first two systems reasonable consistency is found for the models and the experiments, while for the third system too many unquantified parameters exist...

  9. Creep Forming of Carbon-Reinforced Ceramic-Matrix Composites

    Science.gov (United States)

    Vaughn, Wallace L.; Scotti, Stephan J.; Ashe, Melissa P.; Connolly, Liz

    2007-01-01

    A set of lecture slides describes an investigation of creep forming as a means of imparting desired curvatures to initially flat stock plates of carbon-reinforced ceramic-matrix composite (C-CMC) materials. The investigation is apparently part of a continuing effort to develop improved means of applying small CCMC repair patches to reinforced carbon-carbon leading edges of aerospace vehicles (e.g., space shuttles) prior to re-entry into the atmosphere of the Earth. According to one of the slides, creep forming would be an intermediate step in a process that would yield a fully densified, finished C-CMC part having a desired size and shape (the other steps would include preliminary machining, finish machining, densification by chemical vapor infiltration, and final coating). The investigation included experiments in which C-CMC disks were creep-formed by heating them to unspecified high temperatures for time intervals of the order of 1 hour while they were clamped into single- and double-curvature graphite molds. The creep-formed disks were coated with an oxidation- protection material, then subjected to arc-jet tests, in which the disks exhibited no deterioration after exposure to high-temperature test conditions lasting 490 seconds.

  10. On the defect structure due to low energy ion bombardment of graphite

    Science.gov (United States)

    Marton, D.; Bu, H.; Boyd, K. J.; Todorov, S. S.; Al-Bayati, A. H.; Rabalais, J. W.

    1995-03-01

    Graphite surfaces cleaved perpendicular to the c axis have been irradiated with low doses of Ar + ions at 50 eV kinetic energy and perpendicular incidence. Scanning tunneling micrographs (STM) of these irradiated surfaces exhibited dome-like features as well as point defects. These dome-like features retain undisturbed graphite periodicity. This finding is attributed to the stopping of ions between the first and second graphite sheets. The possibility of doping semiconductors at extremely shallow depths is raised.

  11. Numerical algorithms in secondary creep

    International Nuclear Information System (INIS)

    Feijoo, R.A.; Taroco, E.

    1980-01-01

    The problem of stationary creep is presented as well as its variational formulation, when weak constraints are established, capable of assuring one single solution. A second, so-called elasto-creep problem, is further analysed, together with its variational formulation. It is shown that its stationary solution coincides with that of the stationary creep and the advantages of this formulation with respect to the former one is emphasized. Some numerical applications showing the efficiency of the method propesed are finally presented [pt

  12. Creep analysis of orthotropic shells

    International Nuclear Information System (INIS)

    Mehra, V.K.; Ghosh, A.

    1975-01-01

    A method of creep analysis of orthotropic cylindrical shells subjected to axisymmetric loads has been developed. A general study of creep behaviour of cylindrical shells subjected to a uniform internal pressure has been conducted for a wide range of values of anisotropy coefficients and creep law exponent. Analysis includes determination of stress re-distribution, strain rates, stationary state stresses. Application of reference stress technique has been extended to analysis of shells. (author)

  13. Ferrocene-functionalized graphitic carbon nitride as an enhanced heterogeneous catalyst of Fenton reaction for degradation of Rhodamine B under visible light irradiation.

    Science.gov (United States)

    Lin, Kun-Yi Andrew; Lin, Jyun-Ting

    2017-09-01

    To enhance degradation of Rhodamine B (RhB), a toxic xanthene dye, an iron-doped graphitic carbon nitride (CN) is prepared by establishing a covalent bond (-CN-) bridging ferrocene (Fc) and CN via a Schiff base reaction. The π-conjugation between the aromatic Fc and CN can be much enhanced by the covalent bond, thereby facilitating the bulk-to-surface charge transfer and separation as well as reversible photo-redox reactions during photocatalytic reactions. Thus, the resulting Fc-CN exhibits a much higher catalytic activity than CN to activate hydrogen peroxide (HP) for RhB degradation, because the photocatalytically generated electrons from CN can activate HP and effectively maintain the bivalence state of Fe in Fc, which also induces the activation of HP. The RhB degradation by the Fc-CN activated HP process (Fc-CN-HP) is validated to involve OH • by examining the effect of radical probe agent as well as electron paramagnetic resonance (EPR) spectroscopic analysis. Fc-CN is also proven to activate HP for RhB degradation over multiple times without loss of catalytic activity. Through determining the degradation intermediates, RhB is indeed fully decomposed by Fc-CN-HP into much lower-molecular-weight organic compounds. These features indicate that Fc-functionalization can be an advantageous technique to enhance the catalytic activity of CN for activating HP. The results obtained in this study are essential to further design and utilize Fc-functionalized CN for Fenton-like reactions. The findings shown here, especially the degradation mechanism and pathway, are also quite important for treating xanthene dyes in wastewater. Copyright © 2017 Elsevier Ltd. All rights reserved.

  14. Thermal ratcheting and creep damage

    International Nuclear Information System (INIS)

    Clement, G.; Cousseran, P.; Roche, R.L.

    1983-01-01

    Several proposals have been made to assist adesigners with thermal ratcheting in the creep range, the more known has been made by O'DONNELL and POROWSKY. Unfortunately these methods are not validated by experiments, and they take only inelastic distortion into consideration as creep effects. The aim of the work presented here is to correct these deficiencies - in providing an experimental basis to ratcheting analysis rules in the creep range, - in considering the effect of cyclic straining (like cyclic thermal stresses) on the time to rupture by creep. Experimental tests have been performed on austenitic stainless steel at 650 0 C for the first item. Results of these tests and results available in the open literature have been used to built a practical rule of ratcheting analysis. This rule giving a conservative value of the creep distortion, is based on the concept of effective primary stress which is an amplification of the primary stress really applied. Concerning the second point (time to rupture), it was necessary to obtain real creep rupture and not instability. According to the proposal of Pr LECKIE, tests were performed on specimens made out of copper, and of aluminium alloys at temperatures between 150 0 C and 300 0 C. With such materials creep rupture is obtained without necking. Experimental tests show that cyclic straining reduces the time to creep rupture under load controlled stress. Caution must be given to the designer: cyclic thermal stress can lead to premature creep rupture. (orig./GL)

  15. Creep of high temperature composites

    International Nuclear Information System (INIS)

    Sadananda, K.; Feng, C.R.

    1993-01-01

    High temperature creep deformation of composites is examined. Creep of composites depends on the interplay of many factors. One of the basic issues in the design of the creep resistant composites is the ability to predict their creep behavior from the knowledge of the creep behavior of the individual components. In this report, the existing theoretical models based on continuum mechanics principles are reviewed. These models are evaluated using extensive experimental data on molydisilicide-silicon carbide composites obtained by the authors. The analysis shows that the rule of mixture based on isostrain and isostress provides two limiting bounds wherein all other theoretical predictions fall. For molydisilicide composites, the creep is predominantly governed by the creep of the majority phase, i.e. the matrix with fibers deforming elastically. The role of back stresses both on creep rates and activation energies are shown to be minimum. Kinetics of creep in MoSi 2 is shown to be controlled by the process of dislocation glide with climb involving the diffusion of Mo atoms

  16. Creep failure of a spray drier

    CSIR Research Space (South Africa)

    Carter, P

    1998-06-01

    Full Text Available , and creep. The calculations pointed to creep, and no positive metallurgic or physical evidence was discovered to support any of the hypotheses. However, the compression stresses implied that creep deformation could have occurred without inducing discernible...

  17. Progress in radioactive graphite waste management. Additional information

    International Nuclear Information System (INIS)

    2010-06-01

    , especially in the UK. It is intended that this report which contains the proceedings of the conference should contribute to progress in the management of radioactive graphite worldwide. The report contains a selection of the papers presented on various issues related to dismantling and treating irradiated graphite. In addition, the report contains summaries of the four topical discussions which were held during the conference

  18. Development of metal fuel and study of construction materials (I-IV), Part V, Vol. II, Project of the device for irradiation of metal uranium in the reactor; 2. Construction of the loop for uranium radiation creep testing

    International Nuclear Information System (INIS)

    Mihajlovic, A.; Pavlovic, A.

    1965-11-01

    This volume includes the design description for construction of the loop for testing uranium radiation creep. It covers the following: construction of the loop head, protection closure; system for pressure regulation and uranium temperature regulation; system for recording samples dilatation and temperature. Testing of components and the loop on the whole is described as well as the safety reports

  19. Development of fracture toughness test method for nuclear grade graphite

    Energy Technology Data Exchange (ETDEWEB)

    Chang, C. H.; Lee, J. S.; Cho, H. C.; Kim, D. J.; Lee, D. J. [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2006-02-15

    Because of its high strength and stability at very high temperature, as well as very low thermal neutron absorption cross-section, graphite has been widely used as a structural material in Gas Cooled Reactors (GCR). Recently, many countries are developing the Very High Temperature gas cooled Reactor (VHTR) because of the potentials of hydrogen production, as well as its safety and viable economics. In VHTR, helium gas serves as the primary coolant. Graphite will be used as a reflector, moderator and core structural materials. The life time of graphite is determined from dimensional changes due to neutron irradiation, which closely relates to the changes of crystal structure. The changes of both lattice parameter and crystallite size can be easily measured by X-ray diffraction method. However, due to high cost and long time of neutron irradiation test, ion irradiation test is being performed instead in KAERI. Therefore, it is essential to develop the technique for measurement of ion irradiation damage of nuclear graphite. Fracture toughness of nuclear grade graphite is one of the key properties in the design and development of VHTR. It is important not only to evaluate the various properties of candidate graphite but also to assess the integrity of nuclear grade graphite during operation. Although fracture toughness tests on graphite have been performed in many laboratories, there have been wide variations in values of the calculated fracture toughness, due to the differences in the geometry of specimens and test conditions. Hence, standard test method for nuclear graphite is required to obtain the reliable fracture toughness values. Crack growth behavior of nuclear grade graphite shows rising R-curve which means the increase in crack growth resistance as the crack length increases. Crack bridging and microcracking have been proposed to be the dominant mechanisms of rising R-curve behavior. In this paper, the technique to measure the changes of crystallite size and

  20. Characterization of graphite dust produced by pneumatic lift

    Energy Technology Data Exchange (ETDEWEB)

    Shen, Ke [Guangdong Provincial Key Laboratory of Thermal Management Engineering and Materials, Graduate School at Shenzhen, Tsinghua University, Shenzhen 518055, Guangdong (China); Peng, Wei; Liu, Bing [Institute of Nuclear and New Energy Technology of Tsinghua University, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 100084 (China); Kang, Feiyu [Guangdong Provincial Key Laboratory of Thermal Management Engineering and Materials, Graduate School at Shenzhen, Tsinghua University, Shenzhen 518055, Guangdong (China); Yang, Xiaoyong; Li, Weihua [Institute of Nuclear and New Energy Technology of Tsinghua University, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 100084 (China); Yu, Suyuan, E-mail: suyuan@tsinghua.edu.cn [Center for Combustion Energy, The Key Laboratory for Thermal Science and Power Engineering, Ministry of Educations, Tsinghua University, Beijing 100084 (China)

    2016-08-15

    Highlights: • Generation of graphite dust by pneumatic lift. • Determination of morphology and particle size distribution of graphite dust. • The size of graphite dust in this study is compared to AVR and THTR-300 results. • Graphite dust originates from both filler and binder of the matrix graphite. - Abstract: Graphite dust is an important safety concern of high-temperature gas-cooled reactor (HTR). The graphite dust could adsorb fission products, and the radioactive dust is transported by the coolant gas and deposited on the surface of the primary loop. The simulation of coagulation, aggregation, deposition, and resuspension behavior of graphite dust requires parameters such as particle size distribution and particle shape, but currently very limited data on graphite dust is available. The only data we have are from AVR and THTR-300, however, the AVR result is likely to be prejudiced by the oil ingress. In pebble-bed HTR, graphite dust is generally produced by mechanical abrasion, in particular, by the abrasion of graphite pebbles in the lifting pipe of the fuel handling system. Here we demonstrate the generation and characterization of graphite dust that were produced by pneumatic lift. This graphite dust could substitute the real dust in HTR for characterization. The dust, exhibiting a lamellar morphology, showed a number-weighted average particle size of 2.38 μm and a volume-weighted average size of 14.62 μm. These two sizes were larger than the AVR and THTR results. The discrepancy is possibly due to the irradiation effect and prejudice caused by the oil ingress accident. It is also confirmed by the Raman spectrum that both the filler particle and binder contribute to the dust generation.

  1. Tensile cracks in creeping solids

    International Nuclear Information System (INIS)

    Riedel, H.; Rice, J.R.

    1979-02-01

    The loading parameter determining the stress and strain fields near a crack tip, and thereby the growth of the crack, under creep conditions is discussed. Relevant loading parameters considered are the stress intensity factor K/sub I/, the path-independent integral C*, and the net section stress sigma/sub net/. The material behavior is modelled as elastic-nonlinear viscous where the nonlinear term describes power law creep. At the time t = 0 load is applied to the cracked specimen, and in the first instant the stress distribution is elastic. Subsequently, creep deformation relaxes the initial stress concentration at the crack tip, and creep strains develop rapidly near the crack tip. These processes may be analytically described by self-similar solutions for short times t. Small scale yielding may be defined. In creep problems, this means that elastic strains dominate almost everywhere except in a small creep zone which grows around the crack tip. If crack growth ensues while the creep zone is still small compared with the crack length and the specimen size, the stress intensity factor governs crack growth behavior. If the calculated creep zone becomes larger than the specimen size, the stresses become finally time-independent and the elastic strain rates can be neglected. In this case, the stress field is the same as in the fully-plastic limit of power law hardening plasticity. The loading parameter which determines the near tip fields uniquely is then the path-independent integral C*.K/sub I/ and C* characterize opposite limiting cases. The case applied in a given situation is decided by comparing the creep zone size with the specimen size and the crack length. Besides several methods of estimating the creep zone size, a convenient expression for a characteristic time is derived, which characterizes the transition from small scale yielding to extensive creep of the whole specimen

  2. A graphite nanoeraser

    DEFF Research Database (Denmark)

    Liu, Ze; Bøggild, Peter; Yang, Jia-rui

    2011-01-01

    We present here a method for cleaning intermediate-size (up to 50 nm) contamination from highly oriented pyrolytic graphite and graphene. Electron-beam-induced deposition of carbonaceous material on graphene and graphite surfaces inside a scanning electron microscope, which is difficult to remove...... by conventional techniques, can be removed by direct mechanical wiping using a graphite nanoeraser, thus drastically reducing the amount of contamination. We discuss potential applications of this cleaning procedure....

  3. Creep behavior for advanced polycrystalline SiC fibers

    Energy Technology Data Exchange (ETDEWEB)

    Youngblood, G.E.; Jones, R.H. [Pacific Northwest National Lab., Richland, WA (United States); Kohyama, Akira [Kyoto Univ. (Japan)] [and others

    1997-04-01

    A bend stress relaxation (BSR) test has been utilized to examine irradiation enhanced creep in polycrystalline SiC fibers which are under development for use as fiber reinforcement in SiC/SiC composite. Qualitative, S-shaped 1hr BSR curves were compared for three selected advanced SiC fiber types and standard Nicalon CG fiber. The temperature corresponding to the middle of the S-curve (where the BSR parameter m = 0.5) is a measure of a fiber`s thermal stability as well as it creep resistance. In order of decreasing thermal creep resistance, the measured transition temperatures were Nicalon S (1450{degrees}C), Sylramic (1420{degrees}C), Hi-Nicalon (1230{degrees}C) and Nicalon CG (1110{degrees}C).

  4. Oxidation Resistant Graphite Studies

    Energy Technology Data Exchange (ETDEWEB)

    W. Windes; R. Smith

    2014-07-01

    The Very High Temperature Reactor (VHTR) Graphite Research and Development Program is investigating doped nuclear graphite grades exhibiting oxidation resistance. During a oxygen ingress accident the oxidation rates of the high temperature graphite core region would be extremely high resulting in significant structural damage to the core. Reducing the oxidation rate of the graphite core material would reduce the structural effects and keep the core integrity intact during any air-ingress accident. Oxidation testing of graphite doped with oxidation resistant material is being conducted to determine the extent of oxidation rate reduction. Nuclear grade graphite doped with varying levels of Boron-Carbide (B4C) was oxidized in air at nominal 740°C at 10/90% (air/He) and 100% air. The oxidation rates of the boronated and unboronated graphite grade were compared. With increasing boron-carbide content (up to 6 vol%) the oxidation rate was observed to have a 20 fold reduction from unboronated graphite. Visual inspection and uniformity of oxidation across the surface of the specimens were conducted. Future work to determine the remaining mechanical strength as well as graphite grades with SiC doped material are discussed.

  5. Creep Measurement Video Extensometer

    Science.gov (United States)

    Jaster, Mark; Vickerman, Mary; Padula, Santo, II; Juhas, John

    2011-01-01

    Understanding material behavior under load is critical to the efficient and accurate design of advanced aircraft and spacecraft. Technologies such as the one disclosed here allow accurate creep measurements to be taken automatically, reducing error. The goal was to develop a non-contact, automated system capable of capturing images that could subsequently be processed to obtain the strain characteristics of these materials during deformation, while maintaining adequate resolution to capture the true deformation response of the material. The measurement system comprises a high-resolution digital camera, computer, and software that work collectively to interpret the image.

  6. Theoretical treatment of embrittlement by inert gas under creep conditions

    International Nuclear Information System (INIS)

    Beere, W.

    1980-01-01

    Cavities situated on grain boundaries can grow at high temperature under the action of an applied stress. Grain boundary diffusion, surface diffusion, plastic growth, vacancy source control or geometric creep constraint processes may control the rate of cavity growth. The growth mechanisms are compared for the particular application of an irradiated austenitic stainless steel. The calculated growth rates are compared with out of pile measurements of time to fracture of pre-irradiated steel. The comparison is based on the assumption that bubbles nucleate during the initial part of the post irradiation test. The larger bubbles are suitable cavity nuclei and grow. (author)

  7. Vortex pinning and creep experiments

    International Nuclear Information System (INIS)

    Kes, P.H.

    1991-01-01

    A brief review of basic flux-pinning and flux-creep ingredients and a selection of experimental results on high-temperature-superconductivity compounds is presented. Emphasis is put on recent results and on those properties which are central to the emerging understanding of the flux-pinning and flux-creep mechanisms of these fascinating materials

  8. Nanogranular origin of concrete creep.

    Science.gov (United States)

    Vandamme, Matthieu; Ulm, Franz-Josef

    2009-06-30

    Concrete, the solid that forms at room temperature from mixing Portland cement with water, sand, and aggregates, suffers from time-dependent deformation under load. This creep occurs at a rate that deteriorates the durability and truncates the lifespan of concrete structures. However, despite decades of research, the origin of concrete creep remains unknown. Here, we measure the in situ creep behavior of calcium-silicate-hydrates (C-S-H), the nano-meter sized particles that form the fundamental building block of Portland cement concrete. We show that C-S-H exhibits a logarithmic creep that depends only on the packing of 3 structurally distinct but compositionally similar C-S-H forms: low density, high density, ultra-high density. We demonstrate that the creep rate ( approximately 1/t) is likely due to the rearrangement of nanoscale particles around limit packing densities following the free-volume dynamics theory of granular physics. These findings could lead to a new basis for nanoengineering concrete materials and structures with minimal creep rates monitored by packing density distributions of nanoscale particles, and predicted by nanoscale creep measurements in some minute time, which are as exact as macroscopic creep tests carried out over years.

  9. Creep buckling analysis of shells

    International Nuclear Information System (INIS)

    Stone, C.M.; Nickell, R.E.

    1977-01-01

    The current study was conducted in an effort to determine the degree of conservatism or lack of conservatism in current ASME design rules concerning time-dependent (creep) buckling. In the course of this investigation, certain observations were made concerning the numerical solution of creep buckling problems. It was demonstrated that a nonlinear finite element code could be used to solve the time-dependent buckling problem. A direct method of solution was presented which proved to be computationally efficient and provided answers which agreed very well with available analytical solutions. It was observed that the calculated buckling times could vary widely for small errors in computed displacements. The presence of high creep strain rates contributed to the prediction of early buckling times when calculated during the primary creep stage. The predicted time estimates were found to increase with time until the secondary stage was reached and the estimates approached the critical times predicted without primary creep. It can be concluded, therefore, that for most nuclear piping components, whose primary creep stage is small compared to the secondary stage, the effect of primary creep is negligible and can be omitted from the calculations. In an evaluation of the past and current ASME design rules for time-dependent, load controlled buckling, it was concluded that current use of design load safety factors is not equivalent to a safety factor of ten on service life for low creep exponents

  10. Contribution to the influence of selected alloy elements on the strain cycling and creep behaviour of cast iron with spheroidal graphite at temperatures above 450 C; Beitrag zum Einfluss ausgewaehlter Legierungselemente auf das Dehnwechsel- und Zeitstandverhalten von Gusseisen mit Kugelgraphit bei Temperaturen oberhalb 450 C

    Energy Technology Data Exchange (ETDEWEB)

    Michel, Susanne

    2012-02-15

    In this report for the first time an all-embracing databases was raised about the correlation of microstructure and high temperature mechanical properties concerning spheroidal cast iron at temperatures above 450 C. Its basic concept is a systematic variation of alloying elements and benchmarking fatigue and creep behavior of all created heats as a function of microstructure and alloys.

  11. {sup 36}Cl and {sup 14}C behaviour in UNGG graphite during leaching experiments

    Energy Technology Data Exchange (ETDEWEB)

    Pichon, C.; Guy, C.; Comte, J. [Commissariat a l' Energie Atomique - C.E.A., Laboratoire d' Analyses Radiochimiques et Chimiques (L.A.R.C.) 13108 Saint Paul lez Durance (France)

    2008-07-01

    Graphite has been used as a moderator in Natural Uranium Graphite Gas reactors. Among the radionuclides, the long-lived activation product {sup 36}Cl and {sup 14}C, which are abundant in graphite after irradiation can be the main contributors to the dose during disposal. This paper deals with the first results obtained on irradiated graphite from French G2 reactor. Both leaching and diffusion experiments have been performed in order to understand and quantify the radionuclides behaviour. Chlorine leaching seems to be controlled by diffusion transport through graphite matrix. On the contrary {sup 14}C leaching is very low, probably because after irradiation, the remaining {sup 14}C was produced from {sup 13}C activation in the crystalline structure of graphite. (authors)

  12. IFMIF - Design Study for in Situ Creep Fatigue Tests

    International Nuclear Information System (INIS)

    Gordeev, S.; Heinzel, V.; Simakov, St.; Stratmanns, E.; Vladimirov, P.; Moeslang, A.

    2006-01-01

    While the high flux volume (20-50 dpa/fpy) of the International Fusion Materials Irradiation Facility (IFMIF) is dedicated to the irradiation of ∼ 1100 qualified specimens that will be post irradiation examined after disassembling in dedicated Hot Cells, various in situ experiments are foreseen in the medium flux volume (1-20 dpa/fpy). Of specific importance for structural lifetime assessments of fusion power reactors are instrumented in situ creep-fatigue experiments, as they can simulate realistically a superposition of thermal fatigue or creep fatigue and irradiation with fusion relevant neutrons. Based on former experience with in situ fatigue tests under high energy light ion irradiation, a design study has been performed to evaluate the feasibility of in situ creep fatigue tests in the IFMIF medium flux position. The vertically arranged test module for such experiments consists basically of a frame similar to a universal testing machine, but equipped with three pulling rods, driven by independent step motors, instrumentation systems and specimen cooling systems. Therefore, three creep fatigue specimens may be tested at one time in this apparatus. Each specimen is a hollow tube with coolant flow in the specimen interior to maintain individual specimen temperatures. The recently established IFMIF global 3D geometry model was used together the latest McDeLicious code for the neutral and charged particle transport calculations. These comprehensive neutronics calculations have been performed with a fine special resolution of 0.25 cm 3 , showing among others that the specimens will be irradiated with a homogeneous damage rate of up to 13(∼ 9%) dpa/fpy and a fusion relevant damage to helium ratio of 10-12 appm He/dpa. In addition, damage and gas production rates as well as the heat deposition in structural parts of the test module have been calculated. Despite of the vertical gradients in the nuclear heating, CFD code calculations with STAR-CD revealed very

  13. Thermal ratcheting and creep damage

    International Nuclear Information System (INIS)

    Clement, G.; Cousseran, P.; Roche, R.L.

    1983-08-01

    Creep is a cause of deformation; it may also result in rupture in time. Although LMFBR structures are not heavily loaded, they are subjected to large thermal transients. Can structure lifetime be shortened by such transients. Several proposals have been made to assist adesigners with thermal ratcheting in the creep range. Unfortunately these methods are not validated by experiments, and they take only inelastic distorsion into consideration as creep effects. The aim of the work presented here is to correct these deficiencies in providing an experimental basis to ratcheting analysis rules in the creep range, and in considering the effect of cyclic straining (like cyclic thermal stresses) on the time to rupture by creep. Experimental tests have been performed on austenitic stainless steel at 650 0 C for the first item. Results of these tests and results available in the open literature have been used to built a practical rule of ratcheting analysis. This rule giving a conservative value of the creep distortion, is based on the concept of effective primary stress which is an amplification of the primary stress really applied. Concerning the second point (time to rupture), it was necessary to obtain real creep rupture and not instability. According to the proposal of Pr LECKIE, tests were performed on specimen made out of copper, and of aluminium alloys at temperatures between 150 0 C and 300 0 C. With such materials creep rupture is obtained without necking. Experimental tests show that cyclic straining reduces the time to creep rupture under load controlled stress. Caution must be given to the designer: cyclic thermal stress can lead to premature creep rupture

  14. Reassembling Surveillance Creep

    DEFF Research Database (Denmark)

    Bøge, Ask Risom; Lauritsen, Peter

    2017-01-01

    We live in societies in which surveillance technologies are constantly introduced, are transformed, and spread to new practices for new purposes. How and why does this happen? In other words, why does surveillance “creep”? This question has received little attention either in theoretical developm......We live in societies in which surveillance technologies are constantly introduced, are transformed, and spread to new practices for new purposes. How and why does this happen? In other words, why does surveillance “creep”? This question has received little attention either in theoretical...... development or in empirical analyses. Accordingly, this article contributes to this special issue on the usefulness of Actor-Network Theory (ANT) by suggesting that ANT can advance our understanding of ‘surveillance creep’. Based on ANT’s model of translation and a historical study of the Danish DNA database......, we argue that surveillance creep involves reassembling the relations in surveillance networks between heterogeneous actors such as the watchers, the watched, laws, and technologies. Second, surveillance creeps only when these heterogeneous actors are adequately interested and aligned. However...

  15. Softening of metals under hydrogen ion irradiation

    International Nuclear Information System (INIS)

    Guseva, M.I.; Korshunov, S.N.; Martynenko, Yu.V.; Skorlupkin, I.D.

    2005-01-01

    Experimental study results are presented on steel type 18-10 creep under hydrogen ion irradiation. The Irradiation of annealed specimens is accomplished by 15 keV H 2 + ions with a dose up to 10 22 m -2 at current density of 0.6 A/m 2 at temperatures of 570-770 K. Creep tests show that the irradiation at T = 770 K results in a sharp increase of creep rate. At t 570 K the effect of ion-induced creep in steel 18-10 is not observed. The model is proposed which explains the ion-induced creep by accumulation of hydrogen along grain boundaries, their weakening and removal of obstacles to sliding [ru

  16. Temperature-dependence of creep behaviour of dental resin-composites.

    Science.gov (United States)

    El-Safty, S; Silikas, N; Watts, D C

    2013-04-01

    To determine the effect of temperature, over a clinically relevant range, on the creep behaviour of a set of conventional and flowable resin-composites including two subgroups having the same resin matrix and varied filler loading. Eight dental resin-composites: four flowable and four conventional were investigated. Stainless steel split moulds (4 mm × 6 mm) were used to prepare cylindrical specimens for creep examination. Specimens were irradiated in the moulds in layers of 2mm thickness (40s each), as well as from the radial direction after removal from the moulds, using a light-curing unit with irradiance of 650 mW/cm(2). A total of 15 specimens from each material were prepared and divided into three groups (n=5) according to the temperature; Group I: (23°C), Group II: (37°C) and Group III: (45°C). Each specimen was loaded (20 MPa) for 2h and unloaded for 2h. Creep was measured continuously over the loading and unloading periods. At higher temperatures greater creep and permanent set were recorded. The lowest mean creep occurred with GS and GH resin-composites. Percentage of creep recovery decreased at higher temperatures. At 23°C, the materials exhibited comparable creep. At 37°C and 45°C, however, there was a greater variation between materials. For all resin-composites, there was a strong linear correlation with temperature for both creep and permanent set. Creep parameters of resin-composites are sensitive to temperature increase from 23 to 45°C, as can occur intra-orally. For a given resin matrix, creep decreased with higher filler loading. Copyright © 2012 Elsevier Ltd. All rights reserved.

  17. Behavior of ferritic steels irradiated by fast neutrons

    International Nuclear Information System (INIS)

    Erler, Jean; Maillard, Arlette; Brun, Gilbert; Lehmann, Jeanne; Dupouy, J.-M.

    1979-01-01

    Ferritic steels were irradiated in Rapsodie and Phenix at varying doses. The swelling and irradiation creep characteristics are reported below as are the mechanical characteristics of these materials [fr

  18. The behaviour of ferritic steels under fast neutron irradiation

    International Nuclear Information System (INIS)

    Erler, J.; Maillard, A.; Brun, G.; Lehmann, J.; Dupouy, J.M.

    1979-07-01

    Ferritic steels have been irradiated in Rapsodie and Phenix to doses up to 150 dpa F. The swelling and irradiation creep characteristics and the mechanical properties of these materials are reported. (author)

  19. Method for producing dustless graphite spheres from waste graphite fines

    Science.gov (United States)

    Pappano, Peter J [Oak Ridge, TN; Rogers, Michael R [Clinton, TN

    2012-05-08

    A method for producing graphite spheres from graphite fines by charging a quantity of spherical media into a rotatable cylindrical overcoater, charging a quantity of graphite fines into the overcoater thereby forming a first mixture of spherical media and graphite fines, rotating the overcoater at a speed such that the first mixture climbs the wall of the overcoater before rolling back down to the bottom thereby forming a second mixture of spherical media, graphite fines, and graphite spheres, removing the second mixture from the overcoater, sieving the second mixture to separate graphite spheres, charging the first mixture back into the overcoater, charging an additional quantity of graphite fines into the overcoater, adjusting processing parameters like overcoater dimensions, graphite fines charge, overcoater rotation speed, overcoater angle of rotation, and overcoater time of rotation, before repeating the steps until graphite fines are converted to graphite spheres.

  20. Electrochemical treatment of graphite

    Energy Technology Data Exchange (ETDEWEB)

    Podlovilin, V.I.; Egorov, I.M.; Zhernovoj, A.I.

    1983-01-01

    In the course of investigating various modes of electrochemical treatment (ECT) it has been found that graphite anode treatment begins under the ''glow mode''. A behaviour of some marks of graphite with the purpose of ECT technique development in different electrolytes has been tested. Electrolytes have been chosen of three types: highly alkaline (pH 13-14), neutral (pH-Z) and highly acidic (pH 1-2). For the first time parallel to mechanical electroerosion treatment, ECT of graphite and carbon graphite materials previously considered chemically neutral is proposed. ECT of carbon graphite materials has a number of advantages as compared with electroerrosion and mechanical ones with respect to the treatment rate and purity (ronghness) of the surface. A small quantity of sludge (6-8%) under ECT is in highly alkali electrolytes.

  1. Electrochemical treatment of graphite

    International Nuclear Information System (INIS)

    Podlovilin, V.I.; Egorov, I.M.; Zhernovoj, A.I.

    1983-01-01

    In the course of investigating various modes of electroche-- mical treatment (ECT) it has been found that graphite anode treatment begins under the ''glow mode''. A behaviour of some marks of graphite with the purpose of ECT technique development in different electrolytes has been tested. Electrolytes have been chosen of three types: highly alkaline (pH 13-14), neutral (pH-Z) and highly acidic (pH 1-2). For the first time parallel to mechanical electroerosion treatment ECT graphite and carbon graphite materials previously considered chemically neutral is proposed. ECT of carbon graphite materials has a number of advantages as compared with electroerrosion and mechanical ones this is treatment rate and purity (ronghness) of the surface. A sMall quantity of sludge (6-8%) under ECT is in highly alkali electrolytes

  2. Negative creep in nickel base superalloys

    DEFF Research Database (Denmark)

    Dahl, Kristian Vinter; Hald, John

    2004-01-01

    Negative creep describes the time dependent contraction of a material as opposed to the elongation seen for a material experiencing normal creep behavior. Negative creep occurs because of solid state transformations that results in lattice contractions. For most applications negative creep will h...

  3. Mechanisms of transient radiation-induced creep

    International Nuclear Information System (INIS)

    Pyatiletov, Yu.S.

    1981-01-01

    Radiation-induced creep at the transient stage is investigated for metals. The situation, when several possible creep mechanisms operate simultaneously is studied. Among them revealed are those which give the main contribution and determine thereby the creep behaviour. The time dependence of creep rate and its relation to the smelling rate is obtained. The results satisfactorily agree with the available experimental data [ru

  4. Modeling Creep Processes in Aging Polymers

    Science.gov (United States)

    Olali, N. V.; Voitovich, L. V.; Zazimko, N. N.; Malezhik, M. P.

    2016-03-01

    The photoelastic method is generalized to creep in hereditary aging materials. Optical-creep curves and mechanical-creep or optical-relaxation curves are used to interpret fringe patterns. For materials with constant Poisson's ratio, it is sufficient to use mechanical- or optical-creep curves for this purpose

  5. Hydrogen adsorption on and solubility in graphites

    International Nuclear Information System (INIS)

    Kanashenko, S.L.; Wampler, W.R.

    1996-01-01

    The experimental data on adsorption and solubility of hydrogen isotopes in graphite over a wide range of temperatures and pressures are reviewed. Langmuir adsorption isotherms are proposed for the hydrogen-graphite interaction. The entropy and enthalpy of adsorption are estimated, allowing for effects of relaxation of dangling sp 2 bonds. Three kinds of traps are proposed: edge carbon atoms of interstitial loops with an adsorption enthalpy relative to H 2 gas of -4.4 eV/H 2 (unrelaxed, Trap 1), edge carbon atoms at grain surfaces with an adsorption enthalpy of -2.3 eV/H 2 (relaxed, Trap 2), and basal plane adsorption sites with an enthalpy of +2.43 eV/H 2 (Trap 3). The adsorption capacity of different types of graphite depends on the concentration of traps which depends on the crystalline microstructure of the material. The number of potential sites for the 'true solubility' (Trap 3) is assumed to be about one site per carbon atom in all types of graphite, but the endothermic character of this solubility leads to a negligible H inventory compared to the concentration of hydrogen in type 1 and type 2 traps for temperatures and gas pressures used in the experiments. Irradiation with neutrons or carbon atoms increases the concentration of type 1 and type 2 traps from about 20 and 200 appm respectively for unirradiated (POCO AXF-5Q) graphite to about 1500 and 5000 appm, respectively, at damage levels above 1 dpa. (orig.)

  6. Lattice continuum and diffusional creep.

    Science.gov (United States)

    Mesarovic, Sinisa Dj

    2016-04-01

    Diffusional creep is characterized by growth/disappearance of lattice planes at the crystal boundaries that serve as sources/sinks of vacancies, and by diffusion of vacancies. The lattice continuum theory developed here represents a natural and intuitive framework for the analysis of diffusion in crystals and lattice growth/loss at the boundaries. The formulation includes the definition of the Lagrangian reference configuration for the newly created lattice, the transport theorem and the definition of the creep rate tensor for a polycrystal as a piecewise uniform, discontinuous field. The values associated with each crystalline grain are related to the normal diffusional flux at grain boundaries. The governing equations for Nabarro-Herring creep are derived with coupled diffusion and elasticity with compositional eigenstrain. Both, bulk diffusional dissipation and boundary dissipation accompanying vacancy nucleation and absorption, are considered, but the latter is found to be negligible. For periodic arrangements of grains, diffusion formally decouples from elasticity but at the cost of a complicated boundary condition. The equilibrium of deviatorically stressed polycrystals is impossible without inclusion of interface energies. The secondary creep rate estimates correspond to the standard Nabarro-Herring model, and the volumetric creep is small. The initial (primary) creep rate is estimated to be much larger than the secondary creep rate.

  7. Multiaxial creep-fatigue rules

    International Nuclear Information System (INIS)

    Spindler, M.W.; Hales, R.; Ainsworth, R.A.

    1997-01-01

    Within the UK, a comprehensive procedure, called R5, is used to assess the high temperature response of structures. One part of R5 deals with creep-fatigue initiation, and in this paper we describe developments in this part of R5 to cover multiaxial stress states. To assess creep-fatigue, damage is written as the linear sum of fatigue and creep components. Fatigue is assessed using Miner's law with the total endurance split into initiation and growth cycles. Initiation is assessed by entering the curve of initiation cycles vs strain range using a Tresca equivalent strain range. Growth is assessed by entering the curve of growth cycles vs strain range using a Rankine equivalent strain range. The number of allowable cycles is obtained by summing the initiation and growth cycles. In this way the problem of defining an equivalent strain range applicable over a range of endurance is avoided. Creep damage is calculated using ductility exhaustion methods. In this paper we address two aspects; first, the nature of stress relaxation and, hence, accumulated creep strain in multiaxial stress fields; secondly, the effect of multiaxial stress on creep ductility. The effect of multiaxial stress state on creep ductility has been examined using experimental data and mechanistic models. Good agreement is demonstrated between an empirical description of test data and a cavity growth model, provided a simple nucleation criterion is included. A simple scaling factor is applied to uniaxial creep ductility, defined as a function of stress state. The factor is independent of the cavity growth mechanisms and yields a value of equivalent strain which can be conveniently used in determining creep damage by ductility exhaustion. (author). 14 refs, 4 figs

  8. AGC-3 Experiment Irradiation Monitoring Data Qualification Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Laurence Hull

    2014-10-01

    The Graphite Technology Development Program will run a series of six experiments to quantify the effects of irradiation on nuclear grade graphite. The third experiment, Advanced Graphite Creep 3 (AGC 3), began with Advanced Test Reactor (ATR) Cycle 152B on November 27, 2012, and ended with ATR Cycle 155B on April 23, 2014. This report documents qualification of AGC 3 experiment irradiation monitoring data for use by the Very High Temperature Reactor (VHTR) Technology Development Office (TDO) Program for research and development activities required to design and license the first VHTR nuclear plant. Qualified data meet the requirements for data collection and use as described in the experiment planning and quality assurance documents. Failed data do not meet the requirements. Trend data may not meet the requirements, but may still provide some useable information. All thermocouples (TCs) functioned throughout the AGC 3 experiment. There was one interval between December 18, 2012, and December 20, 2012, where 10 NULL values were reported for various TCs. These NULL values were deleted from the Nuclear Data Management and Analysis System database. All temperature data are Qualified for use by the VHTR TDO Program. Argon, helium, and total gas flow data were within expected ranges and are Qualified for use by the VHTR TDO Program. Total gas flow was approximately 50 sccm through the AGC 3 experiment capsule. Helium gas flow was briefly increased to 100 sccm during ATR shutdowns. At the start of the AGC 3 experiment, moisture in the outflow gas line was stuck at a constant value of 335.6174 ppmv for the first cycle (Cycle 152B). When the AGC 3 experiment capsule was reinstalled in ATR for Cycle 154B, a new moisture filter was installed. Moisture data from Cycle 152B are Failed. All moisture data from the final three cycles (Cycles 154B, 155A, and 155B) are Qualified for use by the VHTR TDO Program.

  9. Design project of the experimental device for studying the uranium Creep in the reactor; Predprojekat eksperimentalnog uredjaja za ispitivanje CREEP-a urana u reaktoru

    Energy Technology Data Exchange (ETDEWEB)

    Pavicevic, M [Institute of Nuclear Sciences Boris Kidric, Odeljenje za reaktorsku eksperimentalnu tehniku, Vinca, Beograd (Serbia and Montenegro)

    1963-12-15

    The objective of this task was constructing a device for creep testing of uranium, i.e permanent deformation of the irradiated fuel. Deformation results from the influence of thermal neutron flux, temperature, time of irradiation, mechanical stress. This describes two possible technical solutions experimental device for creep testing in the vertical experimental channel and in the horizontal experimental channel of the RA reactor. In addition to the design details, the report covers calculations of heat generation, antireactivity, activation of the materials of the constructed experimental loop, mechanical calculations as well as description of measurements and regulation of the uranium sample temperature.

  10. Asymptomatic Intracorneal Graphite Deposits following Graphite Pencil Injury

    OpenAIRE

    Philip, Swetha Sara; John, Deepa; John, Sheeja Susan

    2012-01-01

    Reports of graphite pencil lead injuries to the eye are rare. Although graphite is considered to remain inert in the eye, it has been known to cause severe inflammation and damage to ocular structures. We report a case of a 12-year-old girl with intracorneal graphite foreign bodies following a graphite pencil injury.

  11. Creep in electronic ceramics

    Energy Technology Data Exchange (ETDEWEB)

    Routbort, J. L.; Goretta, K. C.; Arellano-Lopez, A. R.

    2000-04-27

    High-temperature creep measurements combined with microstructural investigations can be used to elucidate deformation mechanisms that can be related to the diffusion kinetics and defect chemistry of the minority species. This paper will review the theoretical basis for this correlation and illustrate it with examples from some important electronic ceramics having a perovskite structure. Recent results on BaTiO{sub 3}, (La{sub 1{minus}x}Sr){sub 1{minus}y}MnO{sub 3+{delta}}, YBa{sub 2}Cu{sub 3}O{sub x}, Bi{sub 2}Sr{sub 2}CaCu{sub 2}O{sub x}, (Bi,Pb){sub 2}Sr{sub 2}Ca{sub 2}Cu{sub 3}O{sub x} and Sr(Fe,Co){sub 1.5}O{sub x} will be presented.

  12. Graphite development for gas-cooled reactors in the USA

    International Nuclear Information System (INIS)

    Burchell, T.D.

    1991-01-01

    This document discusses Modular High-Temperature Gas-Cooled Reactor (MHTGR) graphite activities in the USA which currently include the following research and development tasks: coke examination; effects of irradiation; variability of physical properties (mechanical, thermal-physical, and fracture); fatigue behavior, oxidation behavior; NDE techniques; structural design criteria; and carbon-carbon composite control rod clad materials. These tasks support nuclear grade graphite manufacturing technology including nondestructive examination of billets and components. Moreover, data shall be furnished to support design and licensing of graphite components for the MHTGR

  13. Development and engineering plan for graphite spent fuels conditioning program

    International Nuclear Information System (INIS)

    Bendixsen, C.L.; Fillmore, D.L.; Kirkham, R.J.; Lord, D.L.; Phillips, M.B.; Pinto, A.P.; Staiger, M.D.

    1993-09-01

    Irradiated (or spent) graphite fuel stored at the Idaho Chemical Processing Plant (ICPP) includes Fort St. Vrain (FSV) reactor and Peach Bottom reactor spent fuels. Conditioning and disposal of spent graphite fuels presently includes three broad alternatives: (1) direct disposal with minimum fuel packaging or conditioning, (2) mechanical disassembly of spent fuel into high-level waste and low-level waste portions to minimize geologic repository requirements, and (3) waste-volume reduction via burning of bulk graphite and other spent fuel chemical processing of the spent fuel. A multi-year program for the engineering development and demonstration of conditioning processes is described. Program costs, schedules, and facility requirements are estimated

  14. Assessment of management modes for graphite from reactor decommissioning

    International Nuclear Information System (INIS)

    White, I.F.; Smith, G.M.; Saunders, L.J.; Kaye, C.J.; Martin, T.J.; Clarke, G.H.; Wakerley, M.W.

    1984-01-01

    A technological and radiological assessment has been made of the management options for irradiated graphite wastes from the decommissioning of Magnox and advanced gas-cooled reactors. Detailed radionuclide inventories have been estimated, the main contribution being from activation of the graphite and its stable impurities. Three different packaging methods for graphite have been described; each could be used for either sea or land disposal, is logistically feasible and could be achieved at reasonable cost. Leaching tests have been carried out on small samples of irradiated graphite under a variety of conditions including those of the deep ocean bed; the different conditions had little effect on the observed leach rates of radiologically significant radionuclides. Radiological assessments were made of four generic options for disposal of packaged graphite: on the deep ocean bed, in deep geologic repositories at two different types of site, and by shallow land burial. Incineration of graphite was also considered, though this option presents logistical problems. With appropriate precautions during the lifetime of the Cobalt-60 content of the graphite, any of the options considered could give acceptably low doses to individuals, and all would merit further investigation in site-specific contexts

  15. Rupture of Al matrix in U-Mo/Al dispersion fuel by fission induced creep

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Gwan Yoon; Sohn, Dong Seong [UNIST, Daejeon (Korea, Republic of); Kim, Yeon Soo [Argonne National Laboratory, Argonnge (United States); Lee, Kyu Hong [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    This phenomenon was found specifically in the dispersion fuel plate with Si addition in the Al matrix to suppress interaction layer (IL) formation between UMo and Al. It is known that the stresses induced by fission induced swelling in U-Mo fuel particles are relieved by creep deformation of the IL, surrounding the fuel particles, that has a much higher creep rate than the Al matrix. Thus, when IL growth is suppressed, the stress is instead exerted on the Al matrix. The observed rupture in the Al matrix is believed to be caused when the stress exceeded the rupture strength of the Al matrix. In this study, the possibility of creep rupture of the Al matrix between the neighboring U-Mo fuel particles was examined using the ABAQUS finite element analysis (FEA) tool. The predicted rupture time for a plate was much shorter than its irradiation life indicating a rupture during the irradiation. The higher stress leads Al matrix to early creep rupture in this plate for which the Al matrix with lower creep strain rate does not effectively relieve the stress caused by the swelling of the U-Mo fuel particles. For the other plate, no rupture was predicted for the given irradiation condition. The effect of creeping of the continuous phase on the state of stress is significant.

  16. Numerical description of creep of highly creep resistant alloys

    International Nuclear Information System (INIS)

    Preussler, T.

    1991-01-01

    Fatigue tests have been performed with a series of highly creep resistant materials for gas turbines and related applications for gaining better creep data up to long-term behaviour. The investigations were performed with selected individual materials in the area of the main applications down to strains and stresses relevant to design, and have attained trial durations of 25000 to 60000 h. In continuing former research, creep equations for a selection of characterizing individual materials have been improved and partly newly developed on the basis of a differentiated evaluation. Concerning the single materials, there are: one melt each of the materials IN-738 LC, IN-939, IN-100, FSX-414 and Inconel 617. The applied differentiated evaluation is based on the elastoplastical behaviour from the hot-drawing test, the creep behaviour from the non interrupted or the interrupted fatigue test, and the contraction behaviour from the annealing test. The creep equations developed describe the high temperature deformation behaviour taking into account primary, secondary and partly the tertiary creep dependent of temperature, stress and time. These equations are valid for the whole application area of the respective material. (orig./MM) [de

  17. Recent developments in graphite

    International Nuclear Information System (INIS)

    Cunningham, J.E.

    1983-01-01

    Overall, the HTGR graphite situation is in excellent shape. In both of the critical requirements, fuel blocks and support structures, adequate graphites are at hand and improved grades are sufficiently far along in truncation. In the aerospace field, GraphNOL N3M permits vehicle performance with confidence in trajectories unobtainable with any other existing material. For fusion energy applications, no other graphite can simultaneously withstand both extreme thermal shock and neutron damage. Hence, the material promises to create new markets as well as to offer a better candidate material for existing applications

  18. Graphite for fusion energy applications

    International Nuclear Information System (INIS)

    Eatherly, W.P.; Clausing, R.E.; Strehlow, R.A.; Kennedy, C.R.; Mioduszewski, P.K.

    1987-03-01

    Graphite is in widespread and beneficial use in present fusion energy devices. This report reflects the view of graphite materials scientists on using graphite in fusion devices. Graphite properties are discussed with emphasis on application to fusion reactors. This report is intended to be introductory and descriptive and is not intended to serve as a definitive information source

  19. Plasticity and creep of metals

    CERN Document Server

    Rusinko, Andrew

    2011-01-01

    Here is a systematic presentation of the postulates, theorems and principles of mathematical theories of plasticity and creep in metals, and their applications. Special attention is paid to analysis of the advantages and shortcomings of the classical theories.

  20. Creep deformation behaviour and microstructural changes in Zr-2.5% Nb alloy

    International Nuclear Information System (INIS)

    Chaudhuri, S.; Singh, R.; Ghosh, R.N.; Sinha, T.K.; Banerjee, S.

    2002-01-01

    Cold worked and stress relieved Zr-2.5% Nb alloy is a well-known material used as pressure tubes in Pressurised Heavy Water Reactors. The pressure tubes, made of a typical Zr-alloy, consisting of 2.54% Nb, 0.1175% oxygen and less than 100 ppm impurities, are expected to withstand 9.5 MPa to 12.5 MPa pressure at 250 degC to 310 degC under fast neutron fluxes of 3.5 x 10 17 nm -2 s -1 . These tubes are made by hot extrusion at 780 degC with an extrusion ratio 8.3:1 and 40% cold pilgering followed by annealing at 550 degC for 3 hours and subsequently by 20-30% cold pilgering and stress relieving at 400 degC for 24 hours. The microstructure of such cold worked and stress relieved alloy consists of Β-Zr precipitates in the matrix of elongated Α-Zr grains. Although various factors such as irradiation creep, thermal creep, irradiation growth etc are responsible for limiting the life of pressure tubes; the thermal creep contributes significantly in overall creep deformation. Keeping this in view as well as due to non-availability of adequate published information including creep database on this alloy, an extensive investigation on the thermal creep behaviour of indigenously produced Zr-2.5% Nb alloy was undertaken. The creep tests in air using Mayes' creep testing machines were carried out in the temperature range of 300 degC to 450 degC under stresses in the range of 50 to 550 MPa. Analysis of data revealed that the mechanism of creep deformation remains the same in this range

  1. Construction of in-situ creep strain test facility for the SFR fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Gyu; Heo, Hyeong Min; Kim, Jun Hwan; Kim, Sung Ho [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In this study, in-situ laser inspection creep test machine was developed for the measuring the creep strain of SFR fuel cladding materials. Ferritic-martensitic steels are being considered as an attractive candidate material for a fuel cladding of a SFR due to their low expansion coefficients, high thermal conductivities and excellent irradiation resistances to a void swelling. HT9 steel (12CrMoVW) is initially developed as a material for power plants in Europe in the 1960. This steel has experienced to expose up to 200dpa in FFTE and EBR-II. Ferritic-Martensitic steel's maximum creep strength in existence is 180Mpa for 106 hour 600 .deg., but HT9 steel is 60Mpa. Because SFR is difficult to secure in developing and applying materials, HT9 steel has accumulated validated data and is suitable for SFR component. And also, because of its superior dimensional stability against fast neutron irradiation, Ferritic-martensitic steel of 9Cr and 12Cr steels, such as HT9 and FC92(12Cr-2W) are preferable to utilize in the fuel cladding of an SFR in KAERI. The pressurized thermal creep test of HT9 and FC92 claddings are being conducted in KAERI, but the change of creep strain in cladding is not easy to measure during the creep test due to its pressurized and closed conditions. In this paper, in-situ laser inspection pressurized creep test machine developed for SFR fuel cladding specimens is described. Moreover, the creep strain rate of HT9 at 650 .deg. C was examined from the in-situ laser inspection pressurized creep test machine.

  2. Critical survey of the neutron-induced creep behaviour of steel alloys for the fusion reactor materials programme

    International Nuclear Information System (INIS)

    Hausen, H.

    1985-01-01

    The differences between the irradiation environment of a fission reactor and that of a fusion reactor are respectively described in relation to the radiation damage found and expected in the two types of nuclear reactor. It is shown that the microstructure developing for instance in stainless steel alloys is almost invariant to whether the production rate of helium is high or low. The finding is valid up to neutron doses corresponding to about 60 dpa. For this reason, irradiation creep data obtained in fission reactors may be used, with caution, for predicting creep behaviour in fusion reactors.It was further recognized that irradiation creep performed with high energy particles from an accelerator, yields results which are comparable to those obtained in fission reactors. For this reason, simulation creep experiments are found to be valuable for the development of irradiation creep resistant materials using, for example, high energy electrons or protons. Such kind of experiments are performed in many laboratories. For irradiation doses larger than 60 dpa, predictions with respect to creep rates in fission and fusion reactors are difficult. In end-of-life tests, which concern swelling, ductility, tensile properties, rupture, fatigue and embrittlement, the presence of helium, due to its production rate being much higher in most materials exposed to 14 MeV neutrons than to fission neutrons, may be of great importance

  3. Carbon-14 Graphitization Chemistry

    Science.gov (United States)

    Miller, James; Collon, Philippe; Laverne, Jay

    2014-09-01

    Accelerator Mass Spectrometry (AMS) is a process that allows for the analysis of mass of certain materials. It is a powerful process because it results in the ability to separate rare isotopes with very low abundances from a large background, which was previously impossible. Another advantage of AMS is that it only requires very small amounts of material for measurements. An important application of this process is radiocarbon dating because the rare 14C isotopes can be separated from the stable 14N background that is 10 to 13 orders of magnitude larger, and only small amounts of the old and fragile organic samples are necessary for measurement. Our group focuses on this radiocarbon dating through AMS. When performing AMS, the sample needs to be loaded into a cathode at the back of an ion source in order to produce a beam from the material to be analyzed. For carbon samples, the material must first be converted into graphite in order to be loaded into the cathode. My role in the group is to convert the organic substances into graphite. In order to graphitize the samples, a sample is first combusted to form carbon dioxide gas and then purified and reduced into the graphite form. After a couple weeks of research and with the help of various Physics professors, I developed a plan and began to construct the setup necessary to perform the graphitization. Once the apparatus is fully completed, the carbon samples will be graphitized and loaded into the AMS machine for analysis.

  4. Melting temperature of graphite

    International Nuclear Information System (INIS)

    Korobenko, V.N.; Savvatimskiy, A.I.

    2001-01-01

    Full Text: Pulse of electrical current is used for fast heating (∼ 1 μs) of metal and graphite specimens placed in dielectric solid media. Specimen consists of two strips (90 μm in thick) placed together with small gap so they form a black body model. Quasy-monocrystal graphite specimens were used for uniform heating of graphite. Temperature measurements were fulfilled with fast pyrometer and with composite 2-strip black body model up to melting temperature. There were fulfilled experiments with zirconium and tungsten of the same black body construction. Additional temperature measurements of liquid zirconium and liquid tungsten are made. Specific heat capacity (c P ) of liquid zirconium and of liquid tungsten has a common feature in c P diminishing just after melting. It reveals c P diminishing after melting in both cases over the narrow temperature range up to usual values known from steady state measurements. Over the next wide temperature range heat capacity for W (up to 5000 K) and Zr (up to 4100 K) show different dependencies of heat capacity on temperature in liquid state. The experiments confirmed a high quality of 2-strip black body model used for graphite temperature measurements. Melting temperature plateau of tungsten (3690 K) was used for pyrometer calibration area for graphite temperature measurement. As a result, a preliminary value of graphite melting temperature of 4800 K was obtained. (author)

  5. On the anisotropy of in-reactor creep of Zr-2.5Nb tubes

    International Nuclear Information System (INIS)

    Causey, A.R.; Holt, R.A.

    1993-06-01

    Creep specimens made from cold-worked Zr-2.5Nb tubes, fabricated with two different microstructures and crystallographic textures, were irradiated in the Osiris reactor in France in a fast-neutron flux of about 1.8 x 10 18 n.m -2 .s -1 , E > MeV, at 553 and 585 K. The hoop stresses from internal Fluences, up to 4 x 10 25 n.m -2 , more than double those achieved an any other creep test on cold-worked Zr-2.5Nb in which both axial and transverse strain were measured. Creep rates were obtained from strain versus fluence plots, and creep compliances were obtained from plots of the strain rates against hoop stress for each material at each temperature. The ratio of creep rates at 583 K to those at 553 K was ∼ 1.36, a little higher than that extrapolated from stress relaxation results at temperatures between 523 and 568 K. The ratio of the biaxial creep compliance in the axial direction to that in the transverse directions is different for the two test materials: 0.0 to -0.1 for the fuel sheathing texture and 0.5 to 0.6 for the pressure tube texture. The results were analysed using a self-consistent model developed to account for the contributions to the creep anisotropy of the three microstructure parameters involved and to account for the grain interaction effects. The model, which was normalized to test reactor and power reactor creep data for cold-worked Zr-2.5Nb tubes, predicted the ratio of the creep compliancies to be -0.26 and 0.63, respectively. Thus the creep anisotropy of Zr-2.5Nb tubes with pressure-tube-like crystallographic texture can be adequately predicted. (author). 18 refs., 4 tabs., 13 figs

  6. Critical review of creep FRAPCON-3 model under dry storage conditions

    Energy Technology Data Exchange (ETDEWEB)

    Feria, F.; Herranz, L.E. [Unit of Nuclear Safety Research, CIEMAT, Avda. Complutense 22, Madrid, Madrid 28040 (Spain)

    2009-06-15

    There is a general agreement that cladding creep rupture is the most likely and limiting failure mechanism of spent fuel in dry storage compared to other potential mechanisms, like stress corrosion cracking and/or delayed hydride cracking. Nevertheless, occurrence of creep rupture is very improbable since both decay heat and hoop stress tend to decrease throughout dry storage. In spite of this, the current trend to higher burn up levels needs further attention that ensures safe storage of spent fuel irradiated over 45 GWd/MTU. An extensive work has been carried out during the last four decades in the area of in-reactor creep modelling. Unfortunately, the in-reactor conditions are so different from those prevailing under dry storage, that all the experience gained cannot be extrapolated in a straightforward manner. On the other side, as creep tests simulating conditions throughout a 20-40 year dry storage are impractical, post-irradiation cladding creep behaviour has been modelled by means of time-temperature dependent laws developed on the basis of currently available zirconium alloys data. Additionally, some tests have been exploring the effect of irradiation, hydrogen distribution and material composition on the materials creep behaviour. Adaptation of fuel performance codes initially developed for normal and off-normal reactor operation is not an easy task either. Creep modelling is usually dependent of host codes because a good part of its validation and update has been carried out in an integral way, and as a consequence its independent performance assessment is not an easy task. This work examines the current capability of FRAPCON-3 to model creep behaviour under dry storage conditions. To do so, a review of its major fundamentals has been done and its range of applicability discussed. Once its main approximations and drawbacks have been identified, an attempt to overcome some of them has been intended by implementing an alternative expression for creep under

  7. Management of graphite material: a key issue for High Temperature Gas Reactor system (HTGR)

    International Nuclear Information System (INIS)

    Bourdeloie, C.; Marimbeau, P.; Robin, J.C.; Cellier, F.

    2005-01-01

    Graphite material is used in nuclear High Temperature Gas-cooled Reactors (HTGR, Fig.1) as moderator, thermal absorber and also as structural components of the core (Fig.2). This type of reactor was selected by the Generation IV forum as a potential high temperature provider for supplying hydrogen production plants and is under development in France in the frame of the AREVA ANTARES program. In order to select graphite grades to be used in these future reactors, the requirements for mechanical, thermal, physical-chemical properties must match the internal environment of the nuclear core, especially with regard to irradiation effect. Another important aspect that must be addressed early in design is the waste issue. Indeed, it is necessary to reduce the amount of nuclear waste produced by operation of the reactor during its lifetime. Preliminary assessment of the nuclear waste output for an ANTARES type 280 MWe HTGR over 60 year-lifetime gives an estimated 6000 m 3 of activated graphite waste. Thus, reducing the graphite waste production is an important issue for any HTGR system. First, this paper presents a preliminary inventory of graphite waste fluxes coming from a HTGR, in mass and volume, with magnitudes of radiological activities based on activation calculations of graphite during its stay in the core of the reactor. Normalized data corresponding to an output of 1 GWe.year electricity allows comparison of the waste production with other nuclear reactor systems. Second, possible routes to manage irradiated graphite waste are addressed in both the context of French nuclear waste management rules and by comparison to other national regulations. Routes for graphite waste disposal studied in different countries (concerning existing irradiated graphite waste) will be discussed with regard to new issues of large graphite waste from HTGR. Alternative or complementary solutions aiming at lowering volume of graphite waste to be managed will be presented. For example

  8. Assessment of the radiological inventory of EDF's graphite waste through an assimilation method

    International Nuclear Information System (INIS)

    Poncet, B.

    2014-01-01

    The definitive disposal of graphite from the decommissioned UNGG reactors (Chinon A3, Saint-Laurent A1, Saint-Laurent A2 and Bugey 1) has required a radiological inventory of the irradiated graphite. This study focuses on Cl 36 that is produced by neutron absorption on Cl 35 that was present initially in graphite as an impurity (about 80 mg/t of Cl initially in Bugey 1 graphite)). It appears that the changes of Cl 36 concentration along the height of a stack of graphite do neither fit the changes in the neutron flux nor the changes in the graphite temperature. This fact is explained by the high level of purity of the graphite and the nugget effect. Challenged by the absence of spatial correlation of the Cl 36 concentration, an EDF's team has developed an assimilation method based on comparisons between calculations and measurements in order to get a conservative inventory. (A.C.)

  9. Design project of the experimental facility for testing uranium creep in the reactor; Predprojekat eksperimentalnog uredjaja za ispitivanje CREEP-a urana u reaktoru

    Energy Technology Data Exchange (ETDEWEB)

    Pavlovic, A [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1963-12-15

    This report contains the design for constructing the experimental device for testing metal uranium creep in the RA reactor core under defined neutron flux conditions, temperature, mechanical loads and time of irradiation. This device will be placed in one of the experimental channels in the core. This report contains physical, thermal and mechanical calculations and engineering drawings of the device.

  10. Datalogger for the creep laboratory

    International Nuclear Information System (INIS)

    Sambasivan, S.I.; Karthikeyan, T.V.; Chowdhary, D.M.; Anantharaman, P.N.

    1989-01-01

    The creep laboratory, MDL/ICGAR is a facility to study the creep properties of materials which are of interest to the fast reactor programme. The creep test is conducted over a few days to several months and years depending on the test variables employed. In these tests the creep strain and creep rate as a function of time are studied while the load and temperature are kept constant. The datalogger automates the process of recording the strain information as a function of time and also monitors the temperature throughout the test. The system handles 126 temperature channels and 42 strain channels from 27 machines. The temperature inputs are from the thermocouples and for cold junction compensation RTD's are used. An extensometer with a linear variable differential transformer (LVDT) or Super Linear Variable Capacitor (SLVC) form the set up to measure strain. The data logger consists of a front end analog input sub-system (AISS), a 8085 based Data Acquisition System (DAS) communicating to a microcomputer with CP/M operating system. The system responds to the user through the console and outputs of a dot matrix printer. The system, running a real time executive, also allows for on line enabling or disabling of a channel, printing of data, examining the current status and value, setting and getting time etc. (author)

  11. Large earthquakes and creeping faults

    Science.gov (United States)

    Harris, Ruth A.

    2017-01-01

    Faults are ubiquitous throughout the Earth's crust. The majority are silent for decades to centuries, until they suddenly rupture and produce earthquakes. With a focus on shallow continental active-tectonic regions, this paper reviews a subset of faults that have a different behavior. These unusual faults slowly creep for long periods of time and pr